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BWR TRANSIENT ANALYSIS MODEL
WPPSS-FTS-129, Rev. 01
September 1990
Principal Engineers
Y. Y. YUNGS. H. BIAND. E. BUSH
Contributing Engineer
B. M. Moore
Approved: Date:R. O. Vosbur hManager, Saf y '& Reliability Analysis
Date:D. L. LarkinManager, Engineering Analysis & Nuclear Fuel
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DISCLAIMER
This report was prepared by the Washington Public Power Supply
System ("Supply System" ) for submittal to the Nuclear Regulatory
Commission, NRC. The information contained herein is accurate tothe best of the Supply System's knowledge. The use of information
contained in this document by anyone other than the Supply
System, or the NRC is not authorized and with respect to any
unauthorized use, neither the Supply System nor its officers,directors, agents,
responsibility, or
or employees assume any obligation,0
liability or makes any warranty or
representation concerning the contents of this document or itsaccuracy or completeness.
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ACKNOWLEDGEMENTS
The Supply System acknowledges the consulting reviews and
recommendations provided by Energy Incorporated and Yankee Atomic
Electric Company during the course of development of this model.
The Supply System also acknowledges the efforts of Mr. J. C.
Chandler, Consultant, and Mr. Z. . T. Cronin, Yankee Atomic
Electric Company for their reviews and comments on this report.The technical discussions with Philadelphia Electric Company and
Pennsylvania 'Power & Light Company regarding the Peach Bottom0
benchmarks is greatly appreciated.
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ABSTRACT
System-wide models, based on the WNP-2 plant-specific design and
utilizing the RETRAN-02 computer code, are described. Both best-
estimate and conservative models are included. The models are
applicable to a wide range of transients and will.initially be
used for, but not limited to, analysis of the limitingpressurization transients considered for reload core licensing.The best-estimate model is qualified by comparisons to a range ofpower ascension test transients and to the Peach Bottom Unit 2
Turbine Trip Tests. A representative application of the model forlicensing basis calculations of the limiting pressurizationtransients (based on WNP-2 end of cycle 4 conditions) is also
presented.
The benchmark comparisons show good agreement between calculated
and measured data, thereby demonstrating the Supply System's
capability to perform transient analyses for licensingapplications.
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TABLE OF CONTENTS
Pacae
1.0 INTRODUCTION . . . . . . . . . . . . . . . . . . . . 1-1
2.0 MODEL DESCRIPTION 2-1
2.1 Model Geometry
2.1.1 Control Volumes, JunctionsConductors
and Heat
2-6
2-6
2.1.2 Vessel Internals2.1.3 Core Region
2.1.4 Recirculation Loops
2.1.5 Steam and Feedwater Lines
2.2 Component Models
2.2.1 Jet Pumps
2.2.2 Recirculation Pumps
2.2.3 Steam Separators
2,.2.4 Safety/Relief Valves
2.2.5 Core Hydraulics
2.3 Trip Logic
2.4 Control System Models
2.4.1 Feedwater Control System
2.4.2 Pressure Control System
2.4.3 Recirculation Flow Control System
2.4.4 Direct Bypass Heating
2.5 Steady-state Initialization
2-7
2'-8
2-9
2-10
2-16
2-16
2-17
2-17
2-18
2-19
2-21
2-23
2-23
2-24
2-25
2-25
2-35
TABLE OF CONTENTS (Continued)
2.6 RETRAN KineticsPacae
2-35
2.6.1
2.6.2
2.6.3
2.6 '
General Description of the Generationof One-dimensional Data
Calculation of the Initial KineticInput Data
Adjustment of Kinetics Data
Verification of the Supply System'sMethodology
2-36
2-38
2-40
2-41
3.0 QUALIFICATION 3-1
3.1 WNP-2 Power Ascension Tests . .'.* . . . . . . . 3-1
3 ~ 1 ~ 1 Water Level Setpoint Change(Test PAT 23A) 3-4
3.1.1.1 RETRAN Model for Test PAT 23A . 3-4
3.1.1.2 Results and Discussions . . . . 3-4
3 ' ' Pressure Regulator Setpoint Change(Test PAT 22) 3-9
3.1.2.1 RETRAN Model for Test PAT 22 . 3-9
3.1.2.2 Results And Discussions . . . . 3-9
3 ' 3 One Recirculation Pump Trip(Test PAT 30A) 3-15
3.1.3.1 RETRAN Model for Test PAT 30A . 3-15
3.1.3.2 Results And Discussions . . . . 3-16
3.1.4 Generator Load Rejection with Bypass(Test PAT 27) 3-30
3.1.4.1 RETRAN Model for Test PAT 27 . 3-30
3.1.4.2 Results and Discussions . . . . 3-31
TABLE OF CONTENTS (Continued)
3.2 Peach Bottom Turbine Trip Tests
Pacae
3-40
3.2.1 Test Description . . . . . . . . . . . . 3-40
3.2.2 Peach Bottom Unit 2 Model Description . 3-42
3.2.3 Initial Conditions and Model Inputs . . 3-44
3.2.4 Comparison with Measurements . . . . . . 3-47
3.2.4.1 Pressure Comparisons 3-47
3.2.4.2 Power and Reactivity Comparisons 3-62
4.0 LICENSING BASIS ANALYSIS
4.1 Licensing Basis Model
4.2 Load Rejection Without Bypass (LRNB)
4.2.1 Sequence of Events
4.2.2 Results of LRNB RETRAN Analysis
4.3 Feedwater Controller Failure to Maximum( FWCF ) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
4.3.1 Sequence of Events
4.3.2 Results of FWCF RETRAN Analysis
4.4 Summary of Transient Analysis
Demand
4-1
4-2
4-8
4-8
4-10
4-27
4-27
4-28
4-53
5.0 SUMMARY AND CONCLUSIONS . . . . . ~ . ~ . . . . . . 5-1
6.0 REFERENCES . . . . . . . . . . . . ~ ~ ~ . . . . . . 6-1
LIST OF FIGURES
Ficiure
2.1
2.2
2.3
2.4
2 2 1
2.4.1
2;4.2
2.4.3
2.6.1
2. 6.2
2.6.3
Supply System Reload Transient Analysis MethodsComputer Flow Chart
WNP-2 RETRAN Model (Vessel)
WNP-2 RETRAN Model (Active Core Region)
WNP-2 RETRAN Model (Recirculation Loops)
WNP-2 RETRAN Model (Steam Line)
Jet Pump Performance Curve
Feedwater Control System
Pressure Control System
Direct Bypass Heating .
Initial Axial Power DistributionPeach Bottom Unit 2 TT1
Initial Axial Power DistributionPeach Bottom Unit 2 TT2
Initial Axial Power DistributionPeach Bottom Unit 2 TZ3
Pacae
1-5
2-2
2-3
2-4
2-5
2-20
2-27
2-31
2-34
2-43
2-44
2-45
3.F 1
3.1;2
3.1.3
3.1.4
3.1.5
3.1.6
3 1 7
3. 1.8
3. 1.9
3. 1. 10
Feedwater Flow — PAT Test 023
Water Level — PAT Test 023
Dome Pressure - PAT Test 022
Normalized Power — PAT Test 022
Steam Flow — PAT Test 022
Feedwater Flow — PAT Test 022
Recirc Flow Pump A — PAT Test 030A
Recirc Flow Pump B — PAT Test. 030A
Recirc Flow Pump A — PAT Test 030A
Recirc Flow Pump B — PAT Test 030A
1D
1D
3-7
3-8
3-11
3-12
3-13
3-14
3-18
3-19
3-20
3-21
iv
Ficiure
3.1.11
3 ' '23. 1. 13
3. 1. 14
3. 1. 15
3.1.16
3.1.17
3.-1.18
3.1.19
3.1.20
3.1.21
3 '.223.1.23
3.1.24
3 ~ 2
3 2 ~ 1
3.2.2
3.2.3
3.2.4
3.2 '
3.2.6
3.2.7
3.2.8
LIST OF FIGURES (Continued)
Jet Pump A Flow — PAT Test 030A
Zet Pump B Flow — PAT Test 030A
Zet Pump A Flow — PAT Test 030A — 1D
Zet Pump B Flow — PAT Test 030A — 1D
Power — PAT Test 030A
Power — PAT Test 030A — 1D RETRAN
Core Heat Flux — PAT Test 030A'oreHeat Flux — PAT Test 030A
Power — PAT Test 027
RRC Flow A - PAT Test 027
RRC Flow B — .PAT Test 027,.
Total Core Flow — PAT Test 027
Dome Pressure — PAT Test 027
Steam Flow — PAT Test 027
PB2 RETRAN Model
PB TT1 Turbine Inlet Pressure
PB TT2 Turbine Inlet Pressure
PB TT3 Turbine Inlet Pressure
PB TT1 Steam Dome Pressure
PB TT2 Steam Dome Pressure
PB TT3 Steam Dome Pressure
PB TT1 Upper Plenum Pressure
PB TT2 Upper Plenum Pressure
1D
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Pacae
3-22
3-23
3-24
3-25
3-26
3-27
3-28
3-29
3-34
3-35
3-36
3-37
3-38
3-39
3.-43
3-50
3-51
3-52
3-53
3-54
3-55
3-56
3-57
LIST OF FIGURES (Continued)
Ficiure
3.2.9
3.2.10
PB TT3 Upper Plenum Pressure
PB TTl Upper Plenum Pressure
3 ~ 2. 12
3 ~ 2. 13
3.2. 14
3.2.15
3.2.16
3.2.17
3.2.18
3.2.19
3.2.20
3 ' '13 ' '23 ' '33.2.24
3.2.25
3 '.263.2 27
3.2.28
3.2.29
3 '.304.2.1
PB TT3 Upper Plenum Pressure
PB TT1 Core Average Power
PB TT2 Core Average Power
PB TT3 Core Average Power
PB TTl Level A Average LPRM
PB TTl Level B Average 'LPRM ;
PB TTl Level C Average LPRM
PB TT1 Level D Average LPRM
PB TT2 Level A Average LPRM
PB TT2 Level B Average LPRM
PB TT2 Level C Average LPRM
PB TT2 Level D Average LPRM .
PB TT3 Lhvel A Average LPRM .
PB TT3 Level B Average LPRM .
PB TT3 Level C Average LPRM .
PB TT3 Level D Average LPRM
PB TT1 ReactivityPB TT2 Reactivity .
PB TT3 ReactivityWNP-2 LRNB LBM — Steamline Pressure
3.2.11 PB TT2 Upper Plenum Pressure
Pacae
3-58
3-59
3-60
3-61
3-67
3-68
3-69
3-70
3-71
3-72
3-73
3-74
3-75
3-76
3-77
3-78
3-79
3-80
3-81
3-82
3-83
3-84
4-13
Vi
LIST OF FIGURES (Continued)
Ficiure
4.2.2
4.2.3
4.2.4
4.2.5
4.2.6
4.2.7
4.2 '
4 ' '
4.2.10
4.2.11
4.2.12
4.2.13
4.2.14
4 ~ 3 ~ 1
4.3.2
4.3.3
4.3.4
4 '.54.3.6
4 ' '
4.3 '
4 '.94.3.10
WNP-2 LRNB LBM — Vessel Steam Flow
WNP-2 LRNB LBM — Dome PressureA
WNP-2 LRNB LBM — Pressure (Mid-Core)
WNP-2 LRNB LBM — Pressure (Core Exit)WNP-2 LRNB LBM — Total Reactivity .
WNP-2 LRNB LBM — Core Power
WNP-2 LRNB LBM — Core Average Heat Flux
WNP-2 LRNB LBM — Liquid Level
WNP-2 LRNB LBM — Feedwater Flow
WNP-2 LRNB LBM — Void Fraction (Mid-Core)
WNP-2 LRNB LBM — Void Fraction (Core Exit)WNP-2 LRNB LBM — Recirculation Flow
WNP-2 LRNB LBM — Core Inlet Flow
WNP-2 FWCF LBM — Feedwater Flow .
WNP-2 FWCF LBM — Core Inlet Subcooling
WNP-2 FWCF LBM — Liquid Level
WNP-2 FWCF LBM — Turbine Steam Flow .
WNP-2 FWCF LBM — Turbine Bypass Flow
WNP-2 FWCF LBM — Dome Pressure
WNP-2 FWCF LBM — Total ReactivityWNP-2 FWCF LBM — Core Power
WNP-2 FWCF LBM — Core Average Heat Flux
WNP-2 FWCF LBM - Group 1 SRV Flow
Pacae
4-14
4-15
4-16
4-17
4-18
4-19
4-20
4-21
4-22
4-23
4-24
4-25
4-26
4-31
4-32
4-33
4-34
4-35
4-36
4-37
4-38
4-39
4-40
vii
LIST OF FIGURES (Continued)I
Ficiure
4.3.11
4.3.12
4.3.13
4.3.14
4.3.15
4.3.16
4.3.17
4.3.18
4.3.19
4.3.20
4.3.21
4.3.22
WNP-2 FWCF LBM
WNP-2 FWCF LBM
WNP-2 FWCF LBM
WNP-2 FWCF LBM
Group 2 SRV Flow
Group 3 SRV Flow .
Group 4 SRV Flow
Group 5 SRV Flow
WNP-2 FWCF LBM
WNP-2 FWCF LBM
WNP-2 FWCF LBM
WNP-2 FWCF LBM
WNP-2 FWCF LBM
WNP-2 FWCF LBM
WNP-2 FWCF LBM
Core Inlet Flow
Core Exit Flow
Recirculation Flow
Pressure (Mid-Core)
Pressure (Core Exit)Void Fraction (Mid-Core)
Void Fraction (Core Exit)
WNP-2 FWCF LBM — Vessel Steam Flow
~ ~
Pacae
4-41
4-42
4-43
4-44
4-45
4-46
4-47
4-48
4-49
4-50
4-51
4-52
LIST 'OF TABLES
Table
2.1.1
2 ~ 1 ~ 2
2 ~ 1 ~ 3
2.3.1
2.4.1
3 ~ 2 ~ 1
Description of Trip Logic
Control Input DefinitionPeach Bottom Turbine Trip TestsTest Conditions
2-22
2-26
3-41
Pacae
Volume Geometric Data . . . . . . . . . . . . . 2-12
Junction Geometric Data . . . . . . . . . . . . 2-13
Heat Conductor Geometric Data . . . . . . . . . 2-15
3 ' ' Peach Bottom Turbine Trip TestsInitial Conditions 3-46
3.2 ' Peach Bottom Turbine Trip TestsPressure change (Psi) at Peak of FirstOscillation 3-49
3.2.4
3.2.5
Peach Bottom Turbine Trip TestsSummary of Core Average Peak Neutron Power ..Peach Bottom Turbine Trip TestsTime (Sec) of Peak Neutron Power
3-64
3-64
3.2.6 Peach Bottom Turbine Trip TestsSummary of LPRM Level Neutron Power Peaks 3-65
3 '.74.1
4.2
Peach Bottom Turbine Trip TestsReactivity Summary
Input Parameters and Initial TransientConditions, Comparison of Licensing Basisand Nominal plant Conditions
Technical Specification LimitsMaximum Control Rod Insertion Time to
Position'fter
Deenergization of Pilot Valve Solenoids
3-66
4-3
4-6
4.3 Sequence of Events for LRNB Transient 4-9
4 '
4.5
Sequence of Events for Feedwater ControllerFailure t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
Summary of Pre'ssurization Transient Results
4-30
4-53
ix
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1.0 INTRODUCTION
This report describes and presents qualification results of a
transient analysis model for WNP-2. WNP-2 is a boiling water
reactor using a BWR/5 Nuclear Steam Supply System (NSSS) provided
by General Electric (GE) and operated by the Washington Public
Power Supply System (The Supply System). The transient, analysis
model, developed by the Supply System, uses the RETRAN-02 MOD04
("RETRAN-02" or "RETRAN") computer code . As demonstrated inthis topical report, the transient analysis model shows good
agreement to operational transients using plant data and with
proper input assumptions gives limiting or conservative resultsapplicable to core reload or licensing analyses.
RETRAN-02 is a versatile time dependent computer code developed
by the Electric Power Research Institute (EPRI) to model complex
thermal-hydraulic systems. It is capable of providing realisticresponses to system upset conditions. A user inputs a system
model consisting of control volumes, heat slabs, and a flow path
network; for nuclear reactor systems a kinetics model utilizingeither point or one-dimensional kinetics is available.
The RETRAN-02 computer code was developed by utility sponsorship
through EPRI. The computer code qualification studies were based
on code predictions of various separate effects tests, system
effects experiments, and power reactor startup tests which can be
1-1
found in the RETRAN-02 documentation. The NRC Staff's 'Safety
Evaluation Report (SER) for RETRAN-02 has been issued. The NRC
found "... RETRAN-02 is an acceptable computer program for use inlicensing applications for calculating the transients described
in Chapter 15 ... and other transients and events as appropriate
and necessary for nuclear power plant operation.". The RETRAN-02
limitations discussed in the SER and applicable to reload
licensing analysis will be addressed in the Supply System
Applications Topical Report (to be submitted later).
The Supply System has developed the input for the model presented
in this report, representing the WNP-2 plant, from as-builtdrawings and vendor specifications. The nodalization network was
developed over many years by trial and error, comparison tosimilar plant models, and finally through comparison of model
predictions to actual WNP-2 data. This report provides
qualification of the WNP-2 RETRAN model and of the Supply
System's ability to analyze WNP-2 transient behavior through the
application of RETRAN-02 to the analysis of
1. WNP-2 Power Ascension Tests;
2. Peach Bottom 2 Cycle 2 Turbine Trip Tests; and
3. WNP-2 Licensing Basis Analysis.
The Supply System's reload transient analysis methods are based
on the EPRX code package as depicted in Figure 1.1. The steady
1-2
state core physics. codes and models used to provide input to the
transient analysis models are described and qualifiedelsewhere . The SIMTRAN-E MOD3A ("SIMTRAN-E") code collapses
the three-dimensional neutronics data generated by the steady
state core physics codes to the one-dimensional neutronics inputrequired by RETRAN-02 and calculates the moderator density and
fuel temperature dependencies. The one-dimensional kineticsparameter dependencies generated by SIMTRAN-E are modified as
described in Section 2.6 to account for differences between the
RETRAN-02 one-dimensional and SIMULATE-E three-dimensional
moderator 'density calculations. RETRAN-02 is used to model the
NSSS and the VIPRE-01 MOD02 ("VIPRE-01") code is used to model
a single fuel assembly for thermal margin evaluations. Thermal
margin evaluation for WNP-2 is described and qualified in a
separate Applications Topical Report (to be submitted later).
The base WNP-2 RETRAN-02 model is described in Chapter 2. It isdesigned to serve as a best-estimate, time dependent (transient),systems analysis tool. Intended uses of the best-estimate model
include simulator qualification, evaluating operationaltransients, and evaluating proposed system design changes. InChapter 3, the WNP-2 RETRAN-02 model is qualified by comparison
of best-estimate data predictions with plant data collectedduring startup testing. To analyze limiting transients for core
reload design in support of technical specification action, a
Licensing Basis Model is developed by modifying the best estimate
1-3
model with conservative assumptions. The Licensing Basis Model
is described in Chapter 4, which also contains example
calculations with the conservative model. The Applications
Topical Report that will be published later will provide
additional detail on a) the process and interface between the
EPRZ codes for licensing analyses, b) the Licensing Basis Model
input sensitivity and uncertainity analyses, and c) the hotchannel thermal margin methodology. Chapter 5 contains the
Summary and Conclusions and Chapter 6 the References.(
1-4
FlGURE 1.1
Supply System Reload Transient Analysis MethodsComputer Code Flow Chart
CORE PHYSICSANALYSIS
SIMTRAN-E
3-0 to IN Link
MODIFICATIONOFCROSS SECTIONDEPENDENCIES
RETRAN-02
NSSS Model
VIPRE-01
Mot Bundle Model
1-5
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2.0 MODEL DESCRIPTION
This chapter describes the WNP-2 RETRAN-02 Best Estimate Model
developed to analyze a wide range of transients. The Supply
System has been involved with the RETRAN development effort since
1979 and the WNP-2 model has evolved over time with our code
development involvement. In addition, the Supply System has had
two separate and independent reviews performed on the WNP-2
model, by Energy Incorporated (now a Division of NUS Corp) and by
Yankee Atomic. These reviews have provided valuable insightsinto the model and options used and are documented as part of the
model development file.
The best estimate model description is given in the followingsections. Figures 2.1 through 2.4 depict the nodalization scheme
utilized for the WNP-2 model, whereas Section 2.1 provides
additional detail on the Model Geometry, Section 2.2 discusses
models of special components such as the jet pumps and the safetyrelief valves, and the trip logic and control systems models are
described in Sections 2.3 and 2.4, respectively. The steady-
state initialization is discussed in Section 2.5. The WNP-2
model uses twelve active core nodes as is shown in Figure 2.2 and
the interface/input between the RETRAN WNP-2 model and the
reactor physics input to the kinetics model is given in Section
2.6.
2-1
FIGURE 2.1
WNP-2 RETRAN HODEL (Vessel)
Q5
318
28 ga
Qs
pi~
Qs
287
Q9
21
12
(y
18
22 2H
Qi0"
281 288
03
2-2
FIGURE 2 '
WNP-2 RETRAN MODEL (Active Core Region)
CORE OUTLET
62 Qii
61 0Qsi
Is
9
59
58
Qss
Qss
Qs
Qs
Junction No. 8
7
58
Q92
Qs
Heutronlc Region Na 8 55 Qss
5
53
Qss
Qs~
liia,
02 51 0
CORE INLETQi
2-3
s s IS a ~ 00 ~
~i ~
~i !
O
fi ~
~ ~W
To Turbine WNP-2 RETRAN MODEL (Steam Line)
398
Stop388 Valve
Q378
Metwell
SafetyReliefValves
381
382
383
38'1
385
Q3ie
328
O326
ReactorVessel
0363
338
368358
361
358
Inboar dMSIV
BypassLine
2. 1 Model 'Geometry
The physical geometry and volume of the WNP-2 nuclear steam
supply system is modeled as series of control volumes and
junctions. The following sections discuss this physicalarrangement and how it is modeled in the WNP-2 RETRAN model.
2.1.1 Control volumes, Junctions, and Heat Conductors
The control volume nodes are defined as distinct regions withinthe primary system, such as the steam dome or downcomer,'s
shown in Figures 2.1 through 2.4. The physical dimensions of the
volumes, junction areas, lengths, etc. were obtained from WNP-2
plant specific as-built drawings. The nodalization scheme used
has been developed over several years and is based on code
limitations, solution stability, economical run-times, and, most
importantly, agreement to experimental or test data.
In Figures 2.1 through 2.4, the nodal volumes are denoted by
numbers in small circles. The pertinent data associated witheach of these volumes is given in Table 2.1.1. The mass/energy
transfer between volumes is through junctions, depicted inFigures 2.1 through 2.4 as numbered arrows. The junction inputdata is given in Table 2.1.2. In Figure 2.2, the heat conduction
in the active core region is shown as small squares numbered 1
2-6
through 12 for each active core volume. The pertinent data foreach heat conductor is given in Table 2.1.3.
Additional discussion of the salient nodalization by region isgiven in, Subsections 2.1.2 through 2.1.5 below. Features of WNP-2
that are represented by special control volumes, e.g.,safety/relief valves, are discussed in Section 2.2.
2.1.2 Vessel Internals (Refer to Figure 2.1)
Volumes 18, 19, and 20 constitute the middle, lower, and upper
downcomer regions, respectively, of the reactor vessel. The
middle downcomer region, Vol. 18, is the volume in which the
feedwater and the saturated liquid and steam carryunder from the
steam separators are mixed. The upper downcomer region, Vol. 20,
models the liquid/steam interface region and utilizes the RETRAN
"non-equilibrium" option. This option is used in this volume toaccurately account 'or the superheating of steam during
pressurization transients (the pressurization/de-pressurizationeffects are governed by this interface region).
As shown in Figure 2.1, the steam dome, Vol. 16, and the fluidregion below the core support plate (lower plenum), Vol. 3, are
single large volumes. The jet pump volumes, Vol. 1 and 2, are
discussed further in Section 2.2.4.
2-7
The upper plenum region above the upper guide plate, Vol. 13, and
the standpipes, Vol. 14, are both modeled as single volumes. A
single volume, Vol. 15, is used to model the internal region ofthe 225 steam separators. From the steam separator internalregion, the major steam flow is to the steam dome, Vol. 16,
whereas the steam carryunder flows to and mixes in the middle
downcomer, Vol. 18, as mentioned above.
2.1.3 Core Region (Refer to Figure 2.2).
The WNP-2 core region is considered to be the twelve controlvolumes used to model the active region of the core (Vols. 51
through 62), the single volumes for the unheated core inlet and
outlet regions (Vols. 4 and 11), and the core bypass singlevolume region. (The core bypass volume is shown on Figure 2.1 as
Vol. 12) .
There are two means of energy deposition modeled in the core
region. In the active core volumes, twelve heat conductors are
used to represent the reactor fuel. The geometric data for the
heat conductors is presented in Table 2.1.3. In addition, 'a
direct moderator heating model is included to account for directenergy deposition into the active core volumes due to'amma and
neutron heating. For the core bypass volume, a constant fractionof the core thermal power is directly deposited via a RETRAN non-
conducting heat exchanger model.
2-8
As shown in Figure 2.2, the one-dimensional kinetics calculations
are supported by twenty-seven neutronic regions, i.e., twenty-
five in the active core volumes and one per reflector volume.
Thermal-hydraulic feedback of each active core and reflectorvolume is provided to the associated neutronic regions. This
feedback of the calculated water density of each active core and
reflector volume and the attendant corrections are discussed
further in Section 2.6 of this report.
The fuel temperature feedback for each'egion is provided by the
heat conductor models in each active core volume. The average
fuel rod is xepresented by three cylindrical regions — fuel, gap,
and cladding. The cladding is modeled with four nodes, the gap
one node, and the fuel region is modeled with six nodes. The
material conductivity and heat capacity for the U02 fuel and the
Zircaloy cladding are taken from MATPRO and WREM data. For
purposes of the model benchmarking, the gap conductance of the
average core region is assumed to be a constant value
2.1.4 Recirculation Loops (Refer to Figure 2.3)
The two recirculation loops are modeled separately. In each
recirculation loop, five control volumes are used to represent
the recirculation pump and loop piping. In each loop, a singlevolume is used to model ten jet pumps driven by the recirculation
2-9
loop. A special two-stream momentum mixing option is used by
RETRAN to describe the interaction of the recirculation loop
drive flow with the suction flow from the downcomer. Additionaldetail of the jet pump and the recirculation pump modeling isprovided in Section 2.2.
2.1.5 Steam and Feedwater Lines (Refer to Figure 2.4)
The four main steam lines are lumped into one composite line,which is divided into seven control volumes (see Figure 2.4).From the reactor vessel there are three volumes modeling the
steam lines to the Main Steam Isolation Valves (MSIVs), Vols.
310, 320, and 330. The middle of these volumes (Vol. 320) isconnected to the junctions representing the safety/relief valves.
Due to their redundancy, the MSIVs are modeled as a single valve.
The steam lines from the MSIVs to the turbine stop valves are
also modeled in three volumes, Vols. 340, 350, and 360. The
middle of these volumes, Vol. 350, provides a line model for the
turbine bypass, whereas the third volume, Vol. 360 provides the
pressure feedback signal to the Pressure Control System. The
steam line volume between the turbine stop valve and the turbinecontrol valves is modeled as Vol. 390.
The flows from steam line to the turbine (through Zct. 390) and,
subsequently, to the condenser (through Zct. 361) are modeled as
2-10
negative fill junctions with flow rates controlled by the
Pressure Control System (see Section 2.4.2).
The feedwater lines are modeled as a positive fill junction with
flow rate controlled by the Feedwater Control System. For
simulation of very fast transients, the feedwater is constant and
explicit modeling of the lines and pumps is not necessary.
2-11
TABLE 2.1.1
VOLUME GEOMETRIC DATA
FIOW HYDRAULIC
AREA DIAMEKRFT2 (FF
ELEV.
(FF)
123
41112131415
161819
2051"
5253
545556575859606162
201202204205206
207208210211212310320330340350360370390
136.942136.942
2240.00066.640
111.280950.708
. 943,000400.000442.834
6285.3002196.7002498.7001901.700
83.95583.95583.95583.95583.95583.95583.95583.95583.95583.95583.955
125.933148.00030.500
115.00043.50091.710
148.00030.500
115.00043.50091.710
446.430275.370555.400504.280
2747.1601654.540
2.56E+586.750
16.51716.51717.2810.7451.198
14.4433,816
- 8.9186.167
18.54410.22124.1777.8121.0001.0001.0001.0001.0001.0001.0001.0001.0001.0001.0001.500
34.6823.375
21.9791.193
20.11634.6823.375
21.9791.193
20.11633.5092.200
39.0907.1466.861
25.88242.610
2.350
16.51716.51721.4500.7451.198
14.4433.8168.9187.092
21.1008.5319.9607.8121.0001.0001.0001.0001.0001.0001.0001.0001.0001.0001.0001.500
58.36012.02745.3479.727
25.98058.36012.02745.3479.727
25.98037.49023.12546.64142.348
170.589102,74142.61010.001
19.89719.897
114.28089.47483.95565.825
247,12044.85371.807
270.000257.496103.350149.62183.95583.95583.95583.95583.95583.95583.95583.95583.95583.95583.95583.9552.5362.5362.5362.2363.5302.5362.5362.5362.2363.530
11.90811.90811.90811.90816.10416.104
4520.0008.674
1.5921.5920.7810.0450.0450.182
17.7380.5050.641
20.7682.2562.1620.7320.0450.0450.0450.0450.0450.0450.0450.0450.0450.0450.0450.0451.7971.7971.7971.1930.9481.7971.7971.7971.1930.9481.9471.9471.9471.9472.2642.264
75.8622.350
9.9179.9170.000
17.28130.52617.28131.72435.54044.45850.61534.30210.12542.82318.02619.02620.02621.02622.02623.02624.02625.02626.02627.02628.02629.026
-19.409-16.510-15.492
6.4877.680
-19.409-16.510-15.492
6.4877.680
21.46421.206
-15.930-21.792-28.650-40.903
21.460-16.196
JEF PUMP
JEP PUMP
IlNER PIZKi'1(DRE INLET(DRE EKITONE BYPASS
UPPER PLBRMSTANDPIPE
SEPARATOR
MID DOWNCOMER
IQWER DOWNCOtKR
UPPER DOWNKMER
(DRE 51KORE 52KjRE 53
(DRE 54CORE 55(DRE 56GSE 57CORE 58CORE 59GNE 510(DRE 511GNE 512RRC 51 SUCFION
RRC 51 PUMP
RRC 51 HEADER INLETRRC 51 HEADER
RRC 51 RISER
RRC 52 SUCFION
RRC 02 PUMP
RRC 52 HEADER INIETRRC 52 HEADER
RRC 52 RISER
SKAM QlflZFSTEAM LINESKAM LINE
STEAM LINESKAM LINE(QVI3INMEÃFSTEAM LINE
2-12
TABLE 2.1.2
JUNCTION GEOMETRIC DATA
KtmECIS FUNJCl'. VOUJME AREA
NO. FROM 'Io FF2
I
ELEV.
FT
HYDRAULIC
L0SS DIAtEFERKEF. FF
I23
410111213
141516171819
2021
2223
515253
5455
56
575859
6061
201
202204205206207208209
211212213214310320
I 3
2 33 44 51
62 1111 1312 1313 1414 1515 16
19.897019.897022.668054.139063.046483.955055.992045.140033.796030.6800
15 1851 5252 5353 5454 5555 5656 57
57 5858 5959 6060 6161 6219 201
201 202202 204204 205205 206206 I
19 207207 208208 210210 211211 212212 216 310
310 320
51.825083.955083.955083.955083.955083.955083.955083.955083.955083.955083.955083.95502.53602.53601.79242.53603.53000.46092.53602.53601.79242.53603.53000.4609
11.908011.9080
20 16 239.320020 18 149.621018 19 85.1340
4 12 1.50003 12 0.8420
19 I 1.773019 2 1.7730
9.91709.9170
17.281318.026130.526131.724031.724035.540044.458050.625050.625042.833034.302017.281317.281326.434026.434044.503019.026120.026121.026122.026123.026124.026125.026126.026127.026128.026129.026114.3750
-16.5100-15.4920
6.48707.6800
26.434014.3750
-16.5100-15 '920
6.48707.6800
26.434054.000022.4300
0.50890.50890.09800.01010.0161 .
0.01490.11740.10710.43900.08850.06520.04270.06480.11390.20364.73304.73300.06590.01190.01190.01190.01190.01190.01190.01190.01190.01190.01190.0149
11.554413.877511.311911.11575.85503.8530
11.554413.877511.311911.11575.85503.85301.61322.5451
1.83001.8300
-1.00003.26900.41170.73000.68000.4100
-1.0000-1.0000
0.18000.2700
-1.00006.98510.05420.05424.01000.00001.24001.24000.00001,24000.00001.24001.24000.00001.24001.24000.24500.6300
-1.00000.54601.28600.21220.24500.6300
-1.00000.54601.28600.21220.27210.3391
1.59201.59200.19600.02700.03190.04460.30220.50540.15690.41670.05180.73202.75000.00280.00190.21000.21000.08950,04460.04460.04460.04460.04460.04460.04460.04460.04460.04460.04461.79691.79691.79691.79690.94800.10831.79691.79691.79691.79690.94800.10831.94701.9470
JEF PUMP ¹I DIKHJEF PUMP ¹2 DISCH
GNE MZFGNE ¹I MZFKRE ¹12 EXITGNE OmfZFBYPASS OUI1ZFSFANDPIPE MZFSEPARATOR INLEFSEPARA'IOR QJFLZFIlNER DOME MZFMID DOWNGNER 'INLET
IVER DOWNKMER INLEFKRE BYPASS MZF ¹2KRE BYPASS INLEF ¹1JEF PUMP ¹I SUCFION
JEF PUMP ¹2 SUCFION
SEP. m MID IXMKMERKRE ¹I EXITGNE ¹2 EXITGNE ¹3 EXITKRE ¹4 EXITKRE ¹5 EXITKRE ¹6 EXITGNE ¹7 EXITGNE ¹8 EXITGNE ¹9 EGTKRE ¹10 EXITGNE ¹11 EXITRRC LOOP ¹IRRC LOOP ¹1RRC LOOP ¹IRRC UDP ¹IRRC IQOP ¹IRRC 10OP ¹IRRC lOOP ¹2RRC LOOP ¹2RRC KOP ¹2RRC lOOP ¹2RRC LOOP ¹2RRC rmP ¹2SKAM UNESFEAM LINE
2-13
TABLE 2.1.2 (CONT.)
JUNCTION GEOMETRIC DATA
65NECFS FIOW
JCF. VOIHME AREA
NO. FRCN m FF2
HYDRAULIC
EIZV. INDIA IOSS DIAMEKRFF 1 FF COEF. FF DESCRIPFICN
330 320 330 3.6370340 330 340 4.1250350 340 350 . 16.1040360 350 360 16.1040380 360 390 14.1860381 320 370 0.2238382 320 370 0.44773&3 320 370 0.4477384 320 370 0,4477385 320 370 0.4477602 0 13 1.0000601 0 16 1.0000490 0 18 5.0000390 0 390 1.0000
. 361 0 350 1.0000
22.1800-15.1300-21.7920-27.5200-15.0210
21.460021.460021.460021.460021.460031.724069.158041.1000
-16.1960-28.6500
2.92943.73657.07468.48643.76640.97570.97570.97570.97570.97570.00770.03910.01660.57655.2965
0.18520.15410.42031.17802.57620.26300.26300.26300.26300.26300.00000.00000.0000
-1.0000-1.0000
1.07601.14601.94702.26402.12500.37750.37750.37750.37750.37751.12841.12840.13331.12801.3440
SKAM UNESKAM LXNE
SKAM LINESKAM IZKSKAM ICESRV INIEFSRV INIZPSRV INLEFSRV INLEFSRV INLEFHPCS
RCIC LINEFEEDWATER LINE'IURBINE NHG FILLSKAM BYP KEG FILL
2-14
TABLE 2.1.3
HEAT CONDUCTOR GEOMETRIC DATA .
HEAT M)IHME CN: 65DUCIOR SURFACE AREA
(DtD. LEFT RIGHI'EMEIRY VOIlME LEFTRIGIII'O.
INSIDE QJFSIDE) TYPE NO. (FT3 FT2 FT2) DESCRIPTION
12
3
456
78
9
101112
515253
545556
575859
606162
CYL. 1
CYL 1CYL. 1
CYL. 1CYL. 1CYL 1
CYL. 1CYL. 1CYL. 1
CYL 1CYL. 1
CYL. 1
60.2760.2760.2760.2760.2760.2760.2760.2760.2760.2760.2760.27
0.0.0.0.0.0.0.0.0.0.0.0.
5990.5990.5990.5990.5990.5990.5990.5990.5990.5990.5990.5990.
CORE 1
GNE 2GNE 3
GNE 4GNE 5(ORE 6
GNE 7GNE 8G)RE 9 .
GNE 10(DRE 11
GNE 12.
2-15
2.2 Component Models
The transient behavior of a BWR is affected by system components
such as jet pumps, recirculation pumps and steam separators. The
following sections describe the major component models of theWNP-2 RETRAN model.
2.2.1 Jet Pumps
There are ten jet pumps in each of the two recirculation loops.
The ten jet pumps per loop are modeled as a single jet pump. This
jet pump is modeled as a single control volume in the WNP-2
RETRAN model. Jet pump performance is represented by a M-N
characteristic curve (N as a function of M, M = The ratio of the
suction flow to the driving flow, N = The ratio of the specificenergy increase of the suction flow to the specific energy
decrease of the driving flow). To determine RETRAN inputparameters that produce the expected M-N characteristic curve forthe WNP-2 jet pump, a RETRAN sub-model of the recirculation loop
and jet pumps was set up. Pressure distribution data from the
vendor was used to determine the suction and drive nozzle losscoefficients. The momentum mixing formulation was applied to the
suction and nozzle junctions. All other junction and volume
geometry data were calculated using design drawings. The M-N
curve generated with this model is compared to vendor's data in
2-16
Figure 2.2.1. The comparison shows that this modeling technique
provides an acceptable representation for the WNP-2 jet pumps.
2.2.2 Recirculation Pumps
The centrifugal pump model in RETRAN is used to represent theWNP-2 recirculation pumps. The pump unique characteristics (i.e.,moment of inertia, rated values for pump flow, head and torque)
and the pump homologous curves supplied to the RETRAN pump model
are based on pump manufacturer's'ata . Since the recirculationflow control is achieved by varying the position of the flow
control valve, not by varying the pump speed, the recirculationpump motor is modeled with a constant speed.
2.2.3 Steam Separators
The steam separators couple the reactor core and the steam dome.I'he
emphasis in modeling the separators is on achieving the
appropriate coupling between these regions.
The 225 steam separators are modeled as a single control volume.
The Bubble Rise model in RETRAN is used to simulate the separator
performance. Referring to Figure 2.1, the interior of the
separators is represented by Volume 15. The entering two-phase
fluid flow is represented by Junction 14. Separation takes place
2-17
within Volume 15. The exiting steam and liquid flows are
represented by Junction 15 and Junction 23 respectively.
The most important separator input parameters relative to the
coupling between the reactor core and steam dome are the
separator inlet inertia and the pressure drop across the
separators. The separator inertia is determined from vendor's
data . It is calculated as a function of the separator inletquality at the transient initial condition. The separator inletand exit loss coeffic'ients are determined by RETRAN using the
steady state initialization option. The pressure drop
distribution at the rated operating condition is in agreement
with vendor's calculation
2.2.4 Safety/Relief Valves
WNP-2 has 18 relief valves arranged in groups of 2 to 4 valves ata common setpoint. Each of the groups of valves at a common
setpoint is represented by a junction connecting the steam lineto a sink volume in the RETRAN model. The area of the junctionsis taken as the flow area of the valve times the number of valves
being modeled. When the valve is opened with the steam linepressurized, the junction flow becomes choked and the Moody
critical flow option is chosen in RETRAN to calculate the choked
flow rate. Contraction coefficients of the relief valve junctions
2-18
are adjusted to match the specified flow at the reference
pressure.
The opening and closing of the relief valves is modeled using
RETRAN trips. When the pressure in the steam line volume
containing the relief valves reaches the specified setpoint , the
valve is opened linearly after a time delay. The valve stays
fully open until the pressure drops to the reclosing setpoint. Itis then completely closed in a stepwise manner.
2.2.5 Core Hydraulic
The core region is modeled using the compressible, single-stream
flow with momentum flux form of flow equation. The two-phase
friction multiplier is computed with the Baroczy correlation. The
form loss coefficients are set to match values calculated with a
steady-state thermal-hydraulic model. This model was developed
with the FIBWR code and has been benchmarked against plantdata. Initial values of core bypass flow and core support platepressure drop are also determined by steady-state thermal-
hydraulic calculation and input to RETRAN. The algebraic slipmodel allowing the liquid and vapor phases to flow with unequal
velocities is used to simulate the two-phase flow phenomena. The
profile fit subcooled void model is included for neutronic
feedback calculation.
2-19
PXGURE 2.2.1
a
0
JET PUMP PERFORMANCE CURYE
OZ
0Hl-
K
0QGI
Qo
00 0.0 1.0 2.0
FLOH RATIO (H)
.2.3 Trip Logic
The trip capability in RETRAN allows parameters (e.g., elapsed
time, normalized reactor power, pressure, control block output,
etc.) to be tested against a high or low threshold (tripsetpoint). The test can then be used to activate or deactivate a
system. This trip logic is used in the WNP-2 RETRAN model tosimulate the actuation of the Reactor Protection System (RPS) and
to initiate various valves and pump actions. Table 2.3.1
provides a description of the trip logic used in the WNP-2 RETRAN
model. This trip logic can be expanded to incorporate additionaltrips if they are needed.
2-21
TABLE 2.3.1
DESCRIPTION OF TRIP LOGIC
01 End calculation
02 'Aubine Trip(initiate stop valve closure)
03 Initiate MSIV closure
05 Initiate Scram
06 Open S/R valve group 1
-06 reclose S/R valve group 1
CAUSES OF TRIP ACFIVATIOÃ
Simulate transient time > setpoint
Control block -8 (water level) > setpoint (L8)
Contml block -8 (water level) < setpoint (L2)Volte 360 (turbine inlet) pressure < setpoint
Normalized power > setpointVolume 16 (steam dc') pressure > setpointControl block -8 (water level) < setpoint (L3)Trip 502 activatedTrip 503 activate'd
Volume 320 (steam line) pressure > setpoint
Volte 320 (steam line) Pressure < setpoint
Trips K7 thmugh +10 are used for other four S/R valve gmups
11 Trip recirculation pmps
12 Trip FW turbine
13 RCIC initiation
-13 Trip RCIC
14 Initiate HPCS
-14 Trip HPCS
Simulated transient time > setpointTrip 502 activatedVolwe 16 (steam deme) pressure > setpointControl block -8 (water level) < setpoint (L2)
Ccntml block -8 (water level) > setpoint (L8)
Contml block -8 (water level) < setpoint (L2)
Control block -8 (water level) > setpoint (L8)
Contml block -8 (water level) < setpoint (L2)
Contml block -8 (water level) > setpoint (L8)
2-22
2.4 Control System Models
RETRAN control blocks (e.g., integrator, multiplier, function
generator, etc.) provide for the simulation of a variety ofreactor control systems. These control blocks can also be used to
prepare combinations of parameters for editing and to perform
other special functions as may be necessary. All of the RETRAN
minor edit variables aret
inputs. Table 2.4.1 listsavailable for use as control block
the control inputs used in the WNP-2
RETRAN model; The WNP-2 specific control system models are
described below.
2.4.1 Feedwater Control System
The Feedwater Control System comprises a level control system and
a feedwater flow delivery system. The level control system allows
for either one-element or three-element control. In one-element
control, the controller output is only a function of the
difference in setpoint and sensed level. In the three-element
control which is normally used, an additional steam-feed mismatch
is added to the level error. All controller settings and gains
are based on actual plant settings and the vendor's Control
System Design Report . The feedwater delivery system isrepresented by the simulation of the pump flow actuator based on
vendor provided plant specific information.
2-23
Figure 2.4.1 illustrates the WNP-2 Feedwater Control System
model. Upon reactor scram, the Feedwater Control System switches
to one-element control and the water level setpoint is lowered 18
inches.
2.4.2 Pressure Control System
The Pressure Control System is composed of a reactor pressure
regulation system, a turbine control valve system, and a steam
bypass valve system. The signals from the pressure regulationsystem to turbine control valve and steam bypass system can be
regulated either by the difference in turbine inlet pressure and
its setpoint or by the load-speed error signal. The primary
settings which affect the pressure regulation system output are
the regulation gain and lag-lead time constants. They are based
on vendor provided data < . The turbine-generator is not
modeled and the turbine speed is specified as a function oftime.
Figure 2.4.2 illustrates the WNP-2 Pressure Control System model.
Upon a turbine trip, the turbine control valve demand signal isgrounded, thus the turbine bypass valve demand is set ecpxal tothe pressure regulator demand. This will cause the bypass valves
to open immediately, rather than waiting through the pressure
regulator lag time constant.
2-24
2.4.3 Recirculation Flow Control System
WNP-2 is operated with the recirculation flow control system set
in manual control mode. No control element is required and the
flow control valve position is modeled with a function generator.
2.4.4 Direct Bypass Heating
The nonconducting heat exchanger model is used to account fordirect bypass heating. The heat removal rate for this heat
exchanger is determined by a control system. It is assumed to be
a constant fraction of the transient core power as shown inFigure 2.4.3.
2-25
TABLE 2.4.1
CONTROL INPUT DEFINITION
ID VARIABLE
NO. SYMBOL DESCRIPFION
01 WP'*
02 WP*03 LIQV04 LIQV05 LIQV06 KNS07 CONS
08 POWR
09 PRES
10 TRIP11 KNS12 WP"*
13 PRES
18 TIMX19 PRES
21 WQCR
Steam (Jct. 330) flow (8 NBR)
FW (Jct. 490) flow (4 NBR)
Middle downccmer (Vol. 18) liquid volwe (ft**3)lower downcarer (Vol. 19) liquid volme (ft**3)Upper downccmer (Vol. 20) liquid volume (ft*"3)Fraction of total core power deposited directly in core bypass regionConstant of 1.0active core (less core bypass) power (MW)
Turbine inlet (Vol: 390) pressure (psia)Scram (trip ID=5) activation indicatorConstant of 0.0Steam (Jct. 16) flow (8 NBR)
Turbine stop valve inlet (Vol. 360) pressure (psia)Simlation tiae (sec)Turbine bypass inlet (Vol. 350) pressure (psia)Heat transferred frcm clad to coolant for core section 1 (Btu/ibm)
ID No. 22 through 32 are used for heat to coolant for other core sections
50 CCNS
51 TRIP52 63NS
53 KNS
Constant of 1.0Turbine trip (ID=2) activation indicatorTurbine speed reference (100t)load bias (10t)
2-26
FXGURE 2.4.1
Feedwater Control System
jIADC
Va. SUM -014
~ LDC
Vol.
LlqddVol. (fl "3)
SUM -015 FNG -118
Toblo 25
Leva! vs,Volvma
SUM -086 LAG -008
Vol. 12
12
Dryor Flow
lFracllon ol Ralod) MUL -087
G "8.8
Sloam DryarPrassuro Drop
Corroclion
hAUL 0)9L l F
{One ElemonlConlr oil
10Scram
Acllvallon lndlcalor
FXGURE 2.4.1 (CONT.)
Feedwater Control System
SloamI Flow(Fraclion of Rolodl
LAG -OI I
G"100.
g—1.0SUM -007 SUM "009
Filler
LLG -OI0
g-l.O
FW
Flow(Frocllon of Rated)
LAG -006
G"100.
g" 1.0LAG -008 MUL "0l8 Level Feedback
(Three Elemenl Conlral)
50
Cons."1.0
g-l.O SUM "016 g-l.OScrom
Acllvollonhdhcalor
IO
FIGURE 2. 4. 1 (CONT. )
Feedwoter Control System
50 Cans.- 1.0
Lovol Sol Palnlvs. Time
FNG -080Toblo I I
G "563.55
Lovol Sol DownIB'florScram
SUM -082 MUL -08 I
G"- I B.IO
ScramAcllvollonlndicolor
g-1.0
Pl Conlroiler
SUM -004 g"-1.0 SUM -OOI
G "0.0161
INT -002
G"0.024
g-1.0 SUM "003 To FW Pump Flow
Acluolor
MUL -0l9
g» I.B
FXGURE 2.4.1 (CONT.)
Feedwater Control System
FW Pump Flow Actuator
SUM -003 SUM -120 INT -121g-l.O SUM -122 g-l.O g- I.I5
g"-l.O g"-l.2
G" I.25
FNG "126
Toblo l4
FW Flux vs.Acluolor Oulpul
G-I.25
FIGURE 2. 4. 2
Pressure Control System
18Time
Load vs.Time
FNG -012 SUM -454
Load BiosCons.-O. I
g"1.0
52
Turbine SpeedReference
g" l.O SUM -453 g"20. SUM "I IO Speed / Load
Domond ILD)
g"-l.O
18Time FNG "452
Turbine Speedvs. Time
FIGURE 2.4.2 (CONT-)
Time
Turbhe htelPressure '-l.O
Pressure Control System
Pressure Regulator
FNG -013
Pressure Selpointvs Time
g--(Pr es.Re (.-30.) SUM "04)
G-O.03333
LLG -046
g-l.O
LAG -040 DER -049
G-0.274
g" I.O
Presswe
Repvter Demand(PRO)
SUM "053p -I.O LAG -052
G-2.924
INT -05 I SUM -050
g--l.O
Steom Line Compensator
FXGURE .2 (CONT.)
PRD 9Pressure Control System
Byposs Valvel9
Inle I Pr essur e
50
Cons." l.O
9-1.0
SUM -054
9—I.O
3% 8los
FNG -047Table l9G-4.0
LAG -055
Valve Areavs. Poslllon
FNG -033
Tablo 26
MUL -032
G"-992.7 4
ProssuroCorroclion
SUM "450 MUL -451 MIN -027 PRD8ypass Valve Fill Flux
LDTurbine Conlrol Volve Fil Flui
5lg"-I.O
Turbine TripAclivelion hdlco lor
LAG -029 LAG -030 VLM -025
Volvo Aroavs. Poslllon
FNG -035
Toble 27
MUL -034
G--3970.97
PressureCarr ac lion
Conlrol Valve9 Irdul Prossure
FIGURE 2.4.3
Direct Bypass Heating
Acllve Core8 Powor
Cons." I.O
g-l.O
g--1.0
SUM "4IO I DIV -402 LAG -403Sensed Coro
Power
Cons."0.02
Cons;0.02
MUL "404
G"-1.0
ldeol Aemovol Aolo
for Heol Exchongor
2.5 Steady-state Initialization
The RETRAN steady-state initialization option is used toinitialize the model. The parameters specified for the
initialization of WNP-2 model are dome pressure, core inletenthalpy, core flows (flow through core active region and flow
from lower plenum to core inlet region), recirculation flow, jetpump suction flow, feedwater and steam flows.
In addition to the inputs for the thermal-hydraulic
initialization, the values of the various controller setpointsand the initial output of control elements are specified. A nulltransient run is performed to confirm the consistency of the
thermal-hydraulic and control system initialization and to verifythat the steady-state is well established.
2.6 RETRAN Kinetics
The RETRAN-,02 MOD04 code has both point kinetics and one-I
dimensional kinetics capabilities. Selection of point or one-
dimensional kinetics for a given transient depends on the
accuracy requirements of the simulation. Point kinetics is used
in a simulation when the axial power shape is relatively constant
during the period of interest. Pressurization transients are
typically analyzed with one-dimensional kinetics because the
reactivity effects of void collapse and control rod movement play
2-35
an important role in determining the overall results of the
calculation. The one-dimensional kinetics model provides a more
accurate calculation of these effects (particularly control rod
reactivity) than the point kinetics model.
The RETRAN model described herein uses nuclear cross-section
information prepared by the core analysis methodology described
in Reference 2. Files containing kinetics data for the variousfuel types present in the reactor core are produced by CASMO-2
and three-dimensional nodal characteristics of the core are
determined by SIMULATE-E . SIMTRAN-E collapses corewide cross-
sections from three-dimensional form to one-dimensional or pointkinetics form as required by RETRAN. Since SIMULATE-E and RETRAN
calculate moderator density differently, the SIMTRAN-E cross-
sections are adjusted manually to account for thy difference.This process is described more fully in Section 2.6.1 below.
2.6.1 General Description of the Generation of One-dimensional
Data
The first step in the process uses the EPRI codes SIMULATE-E and
SIMTRAN-E. SIMULATE-E predicts core power and burnup
distributions during detailed depletion analyses of the reactorcore. Qualification of the Supply System's SIMULATE-E
methodology is provided elsewhere
2-36
SIMTRAN-E was developed under EPRI sponsorship for linkingSIMULATE-E and RETRAN. SIMTRAN-E reads restart files written by
SIMULATE-E, extracts the appropriate information for determiningcvf'sthe kinetics parameters, and generates the direct, RETRAN input
for transient analysis.
In this first step, SIMTRAN-E produces a one-dimensional kineticsdata file in the form of polynomials which describe the effectsof relative changes in water density and fuel temperature on
calculated two-group cross sections, diffusion coefficients,inverse neutron velocities, radial bucklings, and delayed neutron
fractions. This data file could be used directly by RETRAN.
However, without the adjustments described below, this approach
would lead to very conservative RETRAN predictions for severe
pressurization events. For benchmark analysis of pressurization
events, this conservatism would create artificially large
uncertainty factors.
The kinetics conservatism is due to differences between the
SIMULATE-E model of core average thermal hydraulics and the
RETRAN model of average channel thermal hydraulics. As a resultof these difference, SIMULATE-E and RETRAN predict differentchanges in average moderator density for the same change in core
pressure. Since SIMTRAN-E receives data only from SIMULATE-E, itcannot adjust for the differences. Therefore, a manual
adjustment is made to the SIMTRAN-E output in the second step in
2-37
the kinetics process. This step produces a revised data fileconsisting of a set of adjusted polynomials that can be used
directly by RETRAN in the best estimate mode.
Except as noted in the text, all of the transient benchmark and
example analyses described in this report used the adjusted
kinetics parameters as produced by the second step of the
kinetics process. For transients which do, not involve a
substantial change in moderator density, the adjustment isunnecessary because the induced conservatism is small.
Verification of the Supply System's methodology utilizing a pre-
released version of SIMTRAN-E is discussed in Section 2.6.4,below.
2.6.2. Calculation of the Initial Kinetics Input Data
To begin the first step in the generation of the one-dimensional
data, a nominal SIMULATE-E case is run at a core configurationconsistent with the initial conditions for the given transient.This nominal SIMULATE-E case uses power and void feedback todetermine the three-dimensional core power and flux distributionsand the critical eigenvalue. The nominal SIMULATE-E case uses
cross-sections which have not been adjusted to match k-infinitiesof the lattice physics code (i.e. unadjusted Zal), but it is run
from a SIMULATE-E restart file generated with cross-sections
2-38
which were adjusted to match k-infinities of the lattice physics
code (i.e. adjusted Zal) .
For transients in which it is necessary to model the effects ofcontrol rod insertion resulting from a scram, an additionalSIMULATE-E case is required. This case is based on the nominal
case and is run with power feedback disabled. The only
difference between this case and the nominal case is the controlrod position array, which has all rods fully inserted.
As noted above, SIMTRAN-E reads the restart files generated by
the SIMULATE-E cases. It then collapses the three-dimensional
SIMULATE-E data to one-dimensional data for RETRAN and determines
the dependence of the kinetics parameters on relative water
density, square root of fuel temperature, and control state.
SIMTRAN-E generates all kinetics parameters except kZf> and kZf2
by radial collapse with adjoint flux weighting. kEf] and k~f2
are radially collapsed with volume weighting. The dependence ofthe kinetics parameters on water density and fuel temperature isdetermined by making perturbations in these quantities. Allrelative water density and square root of average fueltemperature perturbations are done in three dimensions and are
then radially collapsed. The one-dimensional nominal and
perturbed kinetics variables are then analyzed to produce
polynomials that are functions of the relative change in water
2-39
I
density and the change in the square root of the average fueltemperature at each axial node. This procedure is performed forboth the nominal case and for the controlled case.
2.6.3. Adjustment of Kinetics Data
To begin the second step in the generation of the, one-dimensional
data, the SIMTRAN-E output from the first step is used and
parallel SIMULATE-E and RETRAN cases are run to quantify the
difference in axial moderator density distributions between the
two models for identical variations in core pressure.
The coefficients in the kinetics parameter polynomials that are
associated with the changes in relative moderator density are
then modified so the change in each kinetics variable in RETRAN
for a given pressure change js the same as that pressure change
would give in the one-dimensional SIMTRAN-E model. Thus
reactivity changes at each axial node generated by a pressure
change are preserved between SIMTRAN-E and RETRAN. The constant
term in each kinetics parameter polynomial determines the initialsteady state eigenvalue in the RETRAN unperturbed state. The
constant terms are not modified when new polynomial fits are
developed; consequently, the SIMULATE-E eigenvalue is preserved
in the RETRAN unperturbed state.
2-40
The cross section libraries used in the core physics analysis are
based on ENDF/B-III. ENDF/B-III includes delayed neutron frac-tions which are artificially low. Preliminary ENDF/B-V data
shows an increase in delayed neutron fraction ranging upwards
from 5.44 in all fissile isotopes. To bring the delayed neutron
fraction closer to those specified in ENDF/B-V, a +54 manual
adjustment is applied to all delayed neutron fractions before
final data is put into the RETRAN input file.
2.6.4. Verification of the Supply System's Methodology
The Supply System's version of SIMTRAN-E is not a formallyreleased EPRI code. Results obtained by this version were
reviewed by EI International and were found to be acceptable
Further verification'as obtained by the Supply System for the
steady state results by comparing the axial power distributionsproduced by SIMTRAN-E for the initial states of the three Peach
Bottom turbine trip tests. Figures 2.6.1, 2.6.2 and 2.6.3 show
the axial power shapes predicted by RETRAN compared to those
predicted by SIMULATE-E and by the process computer for the
initial state of each of these turbine trips. In each case, the
RETRAN model is based on the one-dimensional kinetics inputproduced by SIMTRAN-E and modified as described above.
The ultimate validation of the SIMTRAN-E calculation is the
accuracy with which RETRAN predicts system behavior in benchmark
2-41
transient analyses. The transient mode is validated by the
predictions of the Peach Bottom turbine trip tests, which also
match the data closely. These results are shown in Section 3.2.
2-42
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3.0 QUALIFICATION
The objective of this chapter is to compare the Supply System's
RETRAN simulation with WNP-2 power ascension tests (PAT) and Peach
Bottom turbine trip tests. The Supply System performed these
benchmark analyses to qualify the WNP-2 RETRAN model and todemonstrate user qualifications. The benchmarks comprise fourWNP-2 PAT tests and three Peach Bottom turbine trip tests.
These benchmark analyses, which were performed in the best-estimate
mode, qualify the WNP-2 RETRAN model for the licensing basis
analysis presented in the next chapter.
3.1 WNP-2 Power Ascension Tests
During the period of October -December 1984, a series of power
ascension tests (PAT) at near full power were performed at WNP-2~~.
The data from these tests is available for verifying the WNP-2
RETRAN model. All of the transients analyzed in this chapter were
recorded during the initial WNP-2 PAT testing.
The best-estimate model described in Chapter 2 was used in the PAT
analyses and the best-estimate analyses, discussed below, verifythe modeling. For future reload analysis, the licensing basis model
will be used to assure conservatism in the output. The licensingbasis model is discussed in Chapter 4.
3-1
The power ascension tests chosen for benchmark are as follows:
1. Water level setpoint change —This transient is used mainly tobenchmark the feedwater control system, water level predictionand general stability of the RETRAN model.
2. Pressure regulator setpoint changes — This transient is used
to benchmark the pressure regulator control system, RETRAN
stability and system model accuracy.
3 ~
(
One recirculation pump trip — This transient is to benchmark
the pump coastdown characteristics and system response to an
asymmetric recirculation flow variation.
4. Generator load rejection with bypass — This transient is used
to benchmark the steam line modeling and system pressurizationbehavior.
Since the PAT transients are milder than the limiting transients inlicensing basis analysis, the first three transients were analyzed
using the point-kinetics core modeling. The one-dimensional
kinetics model was also run for the recirculation pump trip case todemonstrate the validity of the point kinetics model for these
relatively mild events.
The load rejection with bypass transient was analyzed using the
3-2
one-dimensional kinetics model. This treatment is consistent withII
the example licensing basis transient analysis (load rejectionwithout bypass) in the next chapter.
3-3
3.1.1 Water level Setpoint'hange (Test PAT 23A)
The purpose of Test PAT 23A was to demonstrate the stability of the
level control system. Test PAT 23A was initiated at 95.14 power
and 96.84 flow. The test procedure consisted of a six-inch step
increase in vessel water level setpoint, a delay to allow the
system to reach a new equilibrium condition, and a six-inch step
decrease in vessel water level setpoint.
3.1.1.1 RETRAN Model for Test PAT 23A
Since the test condition is near the rated condition, the standard
RETRAN base model at rated condition as presented in Chapter 2 isused to start the transient simulation. To model the step change
of the level setpoint, a general function table used in the levelsetpoint control block (Control Block 80) is modified.
3.1.1.2 Results and Discussions
The water level setpoint step change test was analyzed mainly toqualify the feedwater control system and vessel water level models.
It also confirms the adequacy of the RETRAN thermal-hydraulic and
neutronic coupling.
The vessel water level in this test is maintained at a specifiedsetpoint through the variation of the feedwater flow which is
3-4
\
controlled by the feedwater control system. The feedwater
controller uses vessel water level and the mismatch between steam
flow and feedwater flow to determine the feedwater pump speed,
which controls feedwater flow. The controller responds to an
increase in vessel water level setpoint by increasing feedwater
'low. This increased feedwater flow causes water level in downcomer
to increase and the water temperature in core inlet to decrease.
The reduced core inlet temperature leads to a reduced core void,resulting in a core power increase. The feedwater controller,after detecting water level increase, will reduce feedwater flow,which leads to a reduction in power. This combination of feedwater
control logic and core response will lead to a new steady state forboth core power and feedwater flow close to their initial values
and the water level will be stabilized at the new setpoint.
Figure 3.1.1 shows the measured and calculated feedwater flowresponse. Similarly, Figure 3.1.2 shows the measured and calculat-ed water level. These plots show that the RETRAN model predictsevents and timing consistent with the data.
Figure 3.1.2 indicates that RETRAN calculates a water level thatapproaches a value that is six inches higher than the initial water
level at about 20 seconds after the setpoint change. The measured
data indicates a higher asymptotic value of 7.8 inches in water
level change, which may indicate an inconsistency between the levelstep change used in the analysis and actual test.
3-5
Other parameters (steam flow, dome pressure and core power) are not,
plotted because they did not show any significant changes (less
than 34 variation from steady state values) throughout the test.
3-6
Figure 3.1.1
FEEDNATER FLON — PAT TEST 023
RETRAN
PLANT DA A
16
SECONDS24 32
Figure 3.1.2
HATER LEVEL — PAT TEST 023
RETRAN
PLANT DA
NWZoUfZH
08 p i6
SECONOS24 4p
3.1.2 Pressure Regulator Setpoint Change (Test PAT 22)
Test, PAT 22 was performed to demonstrate the stability of the
pressure control system. Test PAT 22 was initiated at 97.54 power
and 95.94 flow. The test procedure consisted of a 10-psi step
decrease in pressure regulator setpoint, a delay to allow the
system to reach a new equilibrium condition, and a 10-psi step
increase in pressure regulator setpoint to the original value.
3.1.2.1 RETRAN Model for Test PAT 22
Test PAT 22 was analyzed with the standard RETRAN base model (cf.Chapter 2). The initial dome pressure in the RETRAN base model is1020 psia, which differs slightly from the 990-psia test pressure.
The transient is very mild and the response to the step change inpressure setpoint was not expected to be sensitive to the small
difference in initial pressure.
To model the test, a general function table used for the pressure
setpoint control block (Control Block 13) is changed to reflect the
step change of the pressure setpoint.
3.1.2.2 Results and Discussions
Zn this test, the pressure regulator and the Digital Electro-Hydraulic Control System (DEH) will respond to a decrease in
3-9
pressure regulator setpoint by opening the turbine control valves
to allow more steam flow which reduces system pressure. The
resulting decreased system pressure increases core voiding and thus
reduces core power. As pressure decreases, the pressure regulatorand DEH control system starts to close the turbine control valve.
The pressure will eventually be stablized at the new setpoint.
Figure 3.1.3 shows the measured and calculated transient pressure
response. The pressure settles out at about 10 psi below the
initial pressure, indicating good alignment of the pressure system
control model.
Figure 3.1.4 presents the measured and calculated power behavior.
The system stabilizes back to the initial power rapidly, and the
RETRAN model predicts this behavior consistently with the data.
Figure 3.1.5 shows the measured and calculated steam flow. Figure
3.1.6 presents the measured and calculated feedwater flow. The
calculation matches the plant data closely in both of these areas.
The simulation/data comparisons indicate that the pressure
regulation control system in the WNP-2 RETRAN model performs as
intended.
3-10
Figure 3.1.3
OOME PRESSURE — PAT TEST 022
+ DOHE PRESSURE
X PLAN'ATA
H0jlL
OJ
8Z
r0 x
x
xx x x x xxxxxxxxx
0I p i2
TAHE (SEC)
Figure 3.1.4
NORMALIZED POWER — PAT TEST 022
+ BETRAN
X PLANT DATA
0lUNHJQoK0Z
XX XX„XX
x x x x xX
80
p i2TrvE (sEc)
Figure 3.1.5
STEAM FLOW — PAT TEST 022
+ FLPM JUN 39P
X PLANT DATA
0Jh.
Q
N ~
H»JXK0Z
X XX
X
Ol
0 p 12
TINE (SEC)24
Figure 3.1.6
FEEDWATER FLOW — PAT TEST 022
+ RETRAN
X PLANT OATA
0JU.
0N ~
HJXK02 X X
80 p f2
TIME (SEC)
3.1.3 One Recirculation Pump Trip (Test PAT 30A)
Test PAT 30A was performed to verify the performance of the
recirculation system. The test also demonstrated that the water
level could be controlled without resulting in turbine trip and/or
scram. Test PAT 30A was initiated at 96.2% power and 1004 flow.
The test was initiated by tripping one recirculation pump.
3.1.3.1 RETRAN Model for Test PAT 30A
Test PAT 30A was analyzed with the standard RETRAN base model atrated power and flow. The transient was initiated by introducing a
recirculation pump trip in Recirculation Loop A at time zero.
A Test PAT 30A case with one-dimensional kinetics was run toevaluate the effect of void feedback on the core power calculationat lower core flow conditions such as in this test and the resultscompared to the point-kinetics model. Unadjusted cross sections
for Beginning of Cycle 1 conditions were used in the one-dimension-
al core analysis. (See Section 2.6 for a description of cross
section adjustments.) Use of the unadj'usted cross sections isacceptable because the one-pump trip transient is very mild. The
data comparison in the next section supports this assumption.
3-15
3.1.3.2 Results and Discussions
The benchmark verifies the coastdown characteristics of the trippedrecirculation pump and the system model response to asymmetric
recirculation flow disturbances. Neutronics, core hydraulics,pressure regulator control system, and feedwater models were
validated in the analysis. Figure 3.1.7 shows measured and
calculated recirculation drive flow for the tripped loop (Loop A)
for the point kinetics case. Figure 3.1.8 shows measured and
calculated recirculation drive flow for the unaffected loop (Loop
B). The calculated flow tracks measured data in both comparisons.
The Loop B flow increases slightly as the transient is initiatedand stabilizes at a higher value. The unaffected loop sees a lower
flow resistance after one pump is tripped.- Figures 3.1.9 and
3.1.10 show the same comparisons for the case using one-dimensional
kinetics. These comparisons are very similar to the cases withpoint-kinetics model, supporting the use of the point kineticsmodel in the other PAT test benchmarks.
Figure 3.1.11 shows the normalized jet pump flow for Loop A.
Figure 3.1.12 shows the jet pump flow for Loop B. Again the RETRAN
results track the data. Figures 3.1.13 and 3.1.14 are the same
comparisons for the case using one-dimensional kinetics. A
comparison of the one-dimensional case with the point-kineticsmodel showed no difference in the calculated jet pump flows.
3-16
Figure 3.1.15 shows that the RETRAN core hydraulic and neutronic
models calculate transient core power consistently with the data.
Initial core power decrease is caused by the increased core void
after the single pump trip. As the core power decreases, the core
void will eventually decrease (after a fuel heat transfer time
constant). The core power begins to increase and level off at a new
equilibrium. Figure 3.1.16 is the corresponding plot for the one-
dimensional RETRAN model. The one-dimensional model gives a
slightly better match with the plant data than the point kineticsmodel later in the transient because the one-dimensional model
tracks the void feedback in the core more accurately than the pointkinetics model. The fluctuations observed at about 4 seconds and.
16 seconds in the one-dimensional case are also the results ofdetailed axial void feedback.
Figures 3.1.17 and 3.1.18 show the core heat flux behavior
calculated by the point-kinetics model and the one-dimensional
model respectively. Both track the plant data with the one-
dimensional model yielding slightly better results.
3-17
Figure 3.1.7
RECIRC FLON PUNP A — PAT TEST 030A
PHP A FLPLANT DA
0Jb.
0hJNHJ<e>oK0z
00 p 12
TIME, SEC24 3p
Figure 3.1.8
RECIRC FLOW PUMP B — PAT TEST 030A
PHP B FL X
PLANT DA A
nUlNHJ<IZ~K0Z
X X X X X X X X X X X X X X
00 0 12
TIME, SECi8 24
Figure 3.1.9
RECIRC FLON PUMP 4 — P4T TEST 0304 — iD
PNP A R.PLANT OA
0tUNHJ<e
K0Z
00
0 i2TINE, SEC
18 24 30
Figure 3.1.10
RECIRC FLON PUMP B — PAT TEST 030A — |D
PHP8FL N
PLANT OA A
0hlNHJ<fm
>oK0Z
X X X X X X X X X X X
O0 12
TINE, SECis 24 30
Figure 3.l.ll
JET PUMP A FLOW — PAT TEST. 030A
FLN JUN 1
PLANT DATA
0IdNHJ<o>oK0Z
XX X X
X X
0I p 12
TINE, SEC
Figure 3.1.12
JET PUMP 8 FLOW — PAT TEST 030A
+ FLON~Qx PLAtP 0ATA
30h.
0llfgN ~
H
XK0z
x x xxxx x x x x x
AOO
TIHE. SEC
Figure 3.1.13
JET PUMP A FLOW — PAT TEST 030A — iD
FLOM 8JN i~ DATA
0blNHJ<p>oK0Z
x x xX X
0I p
TIME, SEC
Figure 3.1.14
JET PUMP B FLOW — PAT TEST 030A — iD
+ FLQI AH Ix PLAUDIT olTA
0>aN ~
JXKOz
x x xxxx xxxxx xxxxx x x
a~o
TIHE, SEC
Figure 3.1.15
POWER — PAT TEST 030A
RHRAN
PLANT DA A
aUJ
NHJQoK0Z
x xx
X X
N
0 0 f2vrvE, sEc
18 24 30
Figure 3.1.16
POWER — PAT TEST 030A — iD RETRAN
RETRAN iPLANT DA A
0rl
II]
0Q.
0IIINHJyOQoK02
xX X
X X
Ol
0 p
TIME, SECi8 24 30
Figure 3. l. 17
CORE HEAT FLUX — PAT TEST 030A
RETRAN
PLANT DATA
X3J0.
>oK0Z
a0 p
TINE, SEC24
Figure 3.1.18
CORE HEAT FLUX — PAT TEST 030A — io
BETRAN 1
PLANT DA A
0rl
X3JtL
~p>oK0z
p0 p 6 12
TINE, SEC3p
3.1.4 Generator Load Rejection With Bypass (Test PAT 27)
Test PAT 27 was initiated at 97.54 power and 95.44 flow. The
transient was started by the activation of the main generator trippushbutton.
The rapid closure of the turbine control valves pressurizes the
steam lines and the core. As the void collapses, positive void
reactivity is induced. Scram is initiated by the turbine controlvalve fast closure pressure switch. The early scram results innegative overall reactivity throughout the test. The net effect isa power decrease shortly after the initiation of the transient. The
turbine control valve closure also initiates the recirculation pump
trip (RPT).
The turbine control valve closure and subsequent steamline
pressurization activate the opening of the turbine bypass valves torelieve vessel pressure. Since the capacity of the bypass is lessthan the test power level, dome pressure continues to increase.
Eventually the SRVs open to limit the pressure rise. For this event
Group 1 SRVs opened.
3.1.4.1 RETRAN Model for Test PAT 27
The main generator trip is set at 0.0 seconds. The turbine controlvalve performance was taken from the test data. In the WNP-2
3-30
RETRAN model a single valve (Junction 380) at the end of steam linesimulates both turbine control and stop valves. When the controlvalve fast closure is activated, its corresponding delay time and
closure time are input so that Junction 380 simulates a controlvalve. .Observed control rod performance data was used as the
RETRAN scram time.
The maximum bypass flow for the base deck is set at the design
value of 254 of rated steam flow. Plant data supports a value ofI
37% maximum bypass flow, which was used for this simulation.
I
The one-dimensional kinetics model was used in this simulation. As
mentioned in Section 3.1.3, for a mild transient as in this case,
uncorrected one-dimensional cross sections are sufficient.
3.1.4.2 Results and Discussions
Figure 3.1.19 shows the calculated and measured variation in the
Average Power Range Monitor (APRM) signal during Test PAT 27. The
APRM signal is proportional to the neutron flux. The output from
RETRAN is adjusted so that the decay power is subtracted from thetotal power before it is compared to the measured data.
Test PAT 27 is the only benchmarked power ascension test which
resulted in a reactor scram. Figure 3.1.19 shows that the RETRAN
prediction tracks the initiation and progress of the scram closely,
3-31
~„~
indicating acceptable scram modeling.
The WNP-2 RETRAN model contains two separate recirculation loops.
Figure 3.1.20 and 3.1.21 show that RETRAN follows the rates ofdecrease for both loops. The lower flow predicted for Loop B is due
to uncertainty of delay time for RPT initiation and a RETRAN
deficiency which results in calculating slightly asymmetrical loop
flows in a symmetric system with symmetric transient conditions.
However, the differences in flows are small. They are not expected
to affect the overall accuracy of the simulation. Figure 3.1.22.
compares the calculated and measured core flow. The results verifythe capability to calculate recirculation loop and core flows aftera Recirculation Pump Trip (RPT).
Figure 3.1.23 shows measured and calculated dome pressure during
the test. Turbine control and stop valve closure causes a rapidsystem pressurization. RETRAN predicts the pressure transientaccurately, particularly during the early stages of the transient(0-4 seconds) when the fuel thermal behavior is important for the
thermal limit predictions. The measured pressure spike at 0.3
seconds appeared only in the wide range Division 2 signal; wide
range Division 1 and narrow range signals do not show thisdeviation. The apparent pressure spike may have been an instrument
aberration. The plant data shows that one relief valve opened
while a second one opened and closed repeatedly. The WNP-2 RETRAN
model treats the first two SRVs with lowest pressure setpoint as a
3-32
single equivalent valve. Both SRVs opened in the RETRAN simulation
and the RETRAN pressure results are lower after about 5 seconds.
Figure 3.1.24 shows the steam flow variation. The oscillation inthe flow rate from 0 to 3 seconds is caused by pressurization waves
after the turbine control and stop valves are closed.
3-33
Figure 3.1.19
POHER — PAT TEST 027
+ POXGl IO NINCTIC
X PLANT CATA
KlU
0Q
0llINMJX„K o
Oo2
00 0 4
TIME, SEC
Figure 3.1.20
RRC FLON A — PAT TEST 027
+ PLlW All ROT -SD
)( PLANT DATA
XX
X
0
030h.
0UJNM
XK0Z
4
00 D 4
TINE. SEC
Figure 3.1.21
RRC FLOH B — PAT TEST 027
+ PLON ~ 404 -lOX PlANT OATA
0h.
0illNHJXK0Z
r4
XX
XX
X
XX
XX
X
00 O 4
TIME, SEC
Figure 3.1.22
TOTAL CORE FLOW — PAT TEST 027
0
0U.
0hlNMJEK0z
'
O
xx x
0O y 4
TIHE, SEC
Figure 3.1.23
OOME PRESSURE — PAT TEST 027
+ PAICS 10 KINETIC
X PLANl'ATA
IUKUlll)IllogaLo
xxxxxx
x
x x x x x xx x x x
004X 4
TIHE, SEC
Figure 3.1.24
STEAM FLOW — PAT TEST 027
+ aa iso n.ox sn
X PLAHz 0ATA
30h.
0IllNHJXK0z
O
X Xx x
X x xXx
X
x
X X X
O
O y 4TIME, SEC
3.2 Peach Bottom Turbine Trip Tests
Benchmarking of the WNP-2 RETRAN model to the power ascension
test measurements verifies the capability and accuracy of most
elements in the model. These benchmarks cover expected operation,but normal startup testing does not cover circumstances which
challenge the core operating limits. To demonstrate the overallcapability of the RETRAN model under design basis conditions, the
Supply System performed an analysis of the three pressurizationtransient tests. (TT1,- TT2, TT3),conducted at Peach Bottom Atomic
Power Station Unit 2 (PB2).
3.2.1 Test Description
Detailed accounts of the Peach Bottom turbine trip tests can be
found in the EPRI report . A brief description is given here.
Turbine trip tests TT1, TT2 and TT3 were performed at PB2 priorto shutdown for refueling at the end of cycle 2 in April 1977.
These tests were conducted jointly by Philadelphia ElectricCompany, GE and EPRI to investigate the effect of the pressure
transient generated in the reactor vessel following a turbinetrip on the neutron flux in the reactor core. Measurements ofmajor process variables such as core pressure, dome pressure and
neutron flux were recorded at a frequency of 166 samples a.
3-40
second. These measurements provide the best set of data for the
qualification of transient analysis methods.
The tests were initiated by manually tripping the turbine
(turbine stop valve closure) at three reactor power levels and
near rated core flow. The turbine stop valve (TSV) position scram
signal was intentionally delayed to allow a significant neutron
flux transient to take place in the core, providing adequate
qualification data. The power increase was terminated by a tripof the reactor protection system on high neutron flux. To
maintain sufficient operating margin, the APRM high neutron fluxscram setpoint was set down closer to the initial reactor power
level. The initial reactor power, core flow and the APRM scram
setpoint for each test are shown in Table 3.2.1.
TABLE 3.2.1
PEACH BOTTOM TURBINE TRIP TESTS
TEST CONDITIONS
POWER CORE FLOW APRM SETPOINT
TEST ~MW Rated Mlbm hr Rated 0 Rated
TT2
TT3
1562
2030
2275
47.4
61.6
69.1
101. 3
82. 9
101.9
98.8
80.9
99.4
85
95
77
3-41
3.2.2 Peach Bottom Unit 2 Model Description
The Peach Bottom model incorporates the modeling techniques ofthe WNP-2 model. A schematic of the model is shown in Figure 3.2.
(The WNP-2 model is shown in Figures 2.1 through 2.4.) The
nodalization within the reactor vessel is identical except thatthe two downcomer volumes are combined into one in the Peach
Bottom model. Also the two recirculation loops are combined intoone in the Peach Bottom model and the recirculation loop isrepresented by two nodes whereas the WNP-2 model has five nodes
for each recirculation loop. The Peach Bottom model includes the
entire main steam bypass system whereas the WNP-2 model uses a
negative fill junction. This model is the best estimate bypass
system model of Hornyik and Naser , and was included to provide
a realistic simulation of this component. Because the steam linegeometry has a significant effect on pressurization transients,the geometric data for the steam line from Philadelphia ElectricCompany's topical report was used. The Peach Bottom steam lineis modeled with six nodes whereas the WNP-2 steam line is modeled
with seven nodes. An additional node was used in the WNP-2 model
to provide more accurate pressure for SRVs lifting. SRVs did not
open during the Peach Bottom turbine trip tests. The physicaldimensions and- characteristics of the dominant fuel type were
used. The dimensions and characteristics for the dominant 7x7
fuel type were obtained from EPRl documentation
3-42
PXGURE 3.2 PB2 RETRAN MODEL
10
19
Feed-waler
2 Loops
13
p; 26,25 j~
5l
30
34
12
14
SRVs
15
Slop valves
Conlrol valves
27 '8 i 29
18
Recirculalion pumps
I
MSIVs
17
Bypassvalves
31
9 Lines
20
Condenser
32 33
3.2.3 Initial Conditions and Model Inputs
Since comparison with actual measurements was a primary
objective, best estimate model inputs developed from originaldata sources such as drawings and measured plant performance data
were used.
The core thermal power, total core flow, core inlet enthalpy and
initial steam flow were obtained from the process computer P-1
edits taken before each transient test. The steam dome pressures
were obtained from the recorded data. The recirculation system
was adjusted to correspond to the initial reactor conditions. The
core bypass flow was calculated for each test with the SIMULATE-E
MOD03 ("SIMULATE-E") computer code . The loss coefficient at the
core bypass region entrance was also adjusted to match the core
plate pressure drop calculated by SIMULATE-E. Reactor water
levels were initialized to match the test data. Table 3.2.2 liststhe initial conditions for each test of the Peach Bottom
analyses.
The feedwater flow rate was specified as a constant value foreach test. The short duration of the tests minimizes the
potential effects of the feedwater control system. The constant
flow assumption was validated through an additional analysisusing feedwater flow characteristics provided by Philadelphia
3-44
Electric Company. Both analyses provided the same results fortransient power and pressure responses.
The control rod scram time and speed can be estimated from the
measured plant Rod Position Information System drift relay output
signals. The average of the measured scram speeds (a sample of31 of the 185 control rods from TT1) is plotted in Re'ference 18
and was used with correction for rod acceleration for all three
tests. All of the control rods were assumed to insert at the
average speed.
The measured time dependent Turbine Stop Valve (TSV) positionand the Turbine Bypass Valve (BPV) position were used as RETRAN
input. A linear TSV opening was assumed with the stroke time
obtained from measured data. The BPV flow area was assumed to be
proportional to the measured position. Since the TSV positionsignal channel malfunctioned during TTl, the average of the
signals from TT2 and TT3 was used.
Since the Peach Bottom Turbine trip tests were pressurizationtransients, they were analyzed using the one-dimensional kineticsmodel. The SIMULATE-E code was used to generate the RETRAN one
dimensional kinetics data at the initial conditions for each
test. A stepwise depletion of cycles 1 and 2 based on the EPRI
documentation was used to determine the fuel exposure, void
history and control history at the time of the tests. The basic
3-45
procedures described in Section 2.6 were used to develop each ofthe three sets of kinetic data.
TABLE 3.2.2
PEACH BOTTOM TURBINE TRIP TESTS
INITIALCONDITIONS
TT2 TT3
Core Thermal Power (MW) 1562 ' 2030.0 275.0
Total Core Flow (ibm/sec)
Core Bypass Flow (ibm/sec)
Core Plate Pressure Drop (psid)
Steam Dome Pressure (psia)
Core Inlet Enthalpy (Btu/ibm)
Steam Flow (ibm/sec)
Recirculation Flow (ibm/sec)
16 '991.6
528.0
1628.0
ll.61
976. 1
518. 1
17. 71
986 '
521.6
2183 ' 2461 '
9386.0 7686.0 9443.0
28139.0 23028.0 28306.0
1636 '0 1384.87 1762.75
3-46
3.2.4 Comparison with Measurements
The goal of these Peach Bottom analyses was to demonstrate the
Supply System's RETRAN model of PB2 can accurately simulate the
power and pressure increase during rapid pressurization
transient. Therefore, comparisons of RETRAN calculations to the
measured pressures (turbine inlet, steam dome and upper plenum)
and LPRM signals are essential. They are presented in the
following two sections.
3.2.4.1 Pressure Comparisons
Pressure comparisons are expressed in terms of pressure change
from the initial value. This eliminates the need to account forthe sensing line elevation head, uncertainties in the pressure
loss coefficient and differences between nodal and localpressures.
Pressure transients calculated at three locations are compared
with the measured values in Figures 3.2.1 through 3.2.3 (at the
turbine inlet), Figures 3.2.4 through 3.2.6 (at steam dome), and
Figures 3.2.7 through 3.2.9 (at core upper plenum). The measured
pressures are obtained directly from the test data file and have
not been. filtered. Time shifts (to account for sensing linedelay) based on information provided in the EPRI documentation
were applied to the calculated pressures. To eliminate the
3-47
instrument line effects, the RETRAN model calculated upper plenum
pressures are compared to the filtered test data in Figures
3.2.10 through 3.2.12.
The good agreements .in turbine inlet pressures and steam dome
pressures indicate that the dynamic of the steam lines is wellsimulated. The first pressure oscillation in the steam dome isslightly overpredicted for TT1 and slightly .underpredicted forTT2 and TT3. The upper plenum pressure calculated by RETRAN forTT1 is slightly higher than. the measured data. For TT2 and TT3,
the calculated and measured upper plenum pressures agree
'reasonably well. Adequate prediction of the core upper plenum
pressure response is essential to transient power predictions.Table 3.2.3 summarizes the calculated and measured steam dome and
upper plenum pressures at the peak of first oscillation.
Overall, the RETRAN calculated pressures are in good agreement
with the measured data. The discrepancies are within the
tolerances due to uncertainties in RETRAN model inputs such as
initial steam flow, bypass valve opening delay and turbine stopvalve closing time.
3-48
TABLE 3.2.3
PEACH BOTTOM TURBINE TRIP TESTS
PRESSURE CHANGE (PSI) AT PEAK OF FIRST OSCILLATION
CALC. DATA
STEAM DOME UPPER PLENUM
CALC. DATA
34.7
40.5
46.2
33.4
41.9
47.4
37.5
42.3
48.3
34.9
44.2
49.9
3-49
FIGURE 3.2.1
PB TTi TURBINE INLET PRESSURE
HEASURED
RETRAN
ld82+OQItI
NWK0.
O
I
0I p
Tlute (sec)
FIGURE 3.2.2
PB TT2 TURBINE INLET PRESSURE
HRog 8
ld8zioOn(
NWKQ.
oOlI
ooI p
TIME (SEC j
FIGURE 3.2.3
PB TT3 TURBINE INLET PRESSURE
HEASURED
RETRAN
IIJ8Z
*aQa
NhlKIL
ONI
O
I p
TINE (SEC)
FIGURE 3-2.4
PB TT1 STEAM DOME PRESSURE
HEASURED
RETRN
OIQ
HNtL
hJ8Z~
z0
NlUKfL
00.
TIME (SEC)
FIGURE 3.2 5
PB TT2 STEAN DONE PRESSURE
}KASUREO
HNQ.
O
td8Z
xD
O~'f
NOlKQ.
O0
TIME (SEC}
FIGURE 3.2.6
PB TT3 STEAN DOME PRESSURE
HEASURED
HNIL
0
Id8Z
z0
O~'f
U)IdKQ.
Oill
O0
TINE (SEC)
PXGURE 3.2.7
PB TTi UPPER PLENUM PRESSURE
MEASURED
BETAAN
06
HNQ.
M8ZQ«f'fx0
NhJKQ.
00
Tree (sec)
FIGURE 3.2.8
OO
PB TT2 UPPER PLENUM PRESSURE
H80.
0
ld8zZ0
O~ V
SlUKIl
Og
O0
nuc (sec)
FIGURE 3.2.9
O0
PB TT3 UPPER PLENUM PRESSURE
HNQ
O
hl82x0
O~'i
SWKQ.
00
TINE fSEC)
FIGURE 3.2.10
PB TTI UPPER PLENUM PRESSURE
MEAS. (FILTERED)
RETRAN
NIvPuKQ.
O0.0 0.6
TINE (SEC)1.2
FIGURE 3.2 11
0l0
PB TT2 UPPER PLENUM PRESSURE
HEAS. tfILTEREO)
RETRAH
HUjo
NUJPu
KIL
D.D 0.6Tzvc (sec)
FIGURE 3.2.12
Oa
PB TT3 UPPER PLENUN PRESSURE
HEAS. (fILTERED)
RETRAH
0.0 0.6Tree (sec)
3.2.4.2 Power an/ Reactivity Comparisons
The calculated core average neutron powers are compared to thenormalized core average LPRM signals for each of the three testsin Figures 3.2.13 through 3.2.15. In all three cases, the rates
of power change, timing trends and power peak magnitudes are ingood agreement with the measured data. The capability to predictthe power transients was further evaluated by comparing the area
under the power peak during the pressurization transients to thatpredicted by RETRAN. In the thermal margin calculation, thecritical power ratio is a function of. the area under the power
peak, not the magnitude of the peak. Tables 3.2.4 presents a
summary of the calculated and measured power peaks and the area
under the peaks for each test. A comparison of the time of the
calculated and measured power peaks is given in Table 3.2.5.
The LPRM detectors are located at four elevations (from bottom totop, levels A, B, C, D). Comparisons of calculated flux tomeasured flux (average of all LPRMs at a level) at each axiallevel are presented in Figures 3.2.16 through 3.2.27. The
calculated and measured peak values at each level is summarized
in Table 3.2.6. Timing trend and relative peak magnitude are
predicted well in the individual LPRM levels. This demonstrates
that RETRAN is capable of predicting accurately the power shape
change during rapid pressurization transient.
3-62
The calculated net. reactivity, scram reactivity, and net
reactivity implied by the data are presented in Figures 3.2.28
through 3.2.30. The implied net reactivity was calculated using
an inverse point kinetic algorithm and the average of the
measured LPRM signals. Table 3.2.7 summarizes the calculated and
implied peak reactivities. The agreement between the calculated
and the implied net reactivities is good. The slightoverprediction of the peak reactivity for all three tests iscaused by the overprediction of the upper plenum pressure at the
time of peak reactivity. It is noted that the magnitudes of the
overpredictions in reactivity are small relative to those of the
neutron power. This is due to the extreme sensitivity of the
neutron power to .changes in reactivity close to the prompt
critical condition. The reactivity figures also show that thecalculated TT1 and TT2 net reactivity turns before scram occurs
while TT3 net reactivity is turned by scram. The same trends are
observed for the implied net reactivities.
3-63
TABLE 3.2.4
PEACH BOTTOM TURBINE TRIP TESTS
SUMMARY OF CORE AVERAGE PEAK NEUTRON 'POWER
PEAK NEUTRON POWER (NORM)
CALC. DATA O'IFF.AREA UNDER PEAK
CALC. DATA DIFF.
5.41'.68
5.39
4.83
4.54
4.90
12.0
3.1
10. 0
0.960
0.769
0 '170.743 3.5
0.669 7.2
0.888 8.1
TABLE 3.2.5
PEACH BOTTOM TURBINE TRIP TESTS
TIME (SEC) OF PEAK NEUTRON POWER
CALC. DATA
TT1
TT2
TT3
.774
.720
.702
.774
.726
.702
3-64
TABLE 3.2 '
PEACH BOTTOM TURBINE TRIP TESTS
SUMMARY OF LPRM LEVEL NEUTRON POWER PEAKS
CORE
D AVG.
Calculation
Data
4 Diff.
3 '23.48
6.90
4.98
4.46
11.'7
5.99
5.23
14.5
6.15
5.59
10.0
5.41
4.83
12.0
Calculation
Data
4 Diff.
3.49
3.52
-0.9
~4. 68
4.50
4.0
5. 09
4 '13 ~ 7
4.82
5.02
—.4.0
4.68
4.54
3.1
Calculation
Data
4 Diff.
3.84
3.68
4.3
5.42
4.83
12.2
6.06
5.45
11.2
5.74
5.47
4.9
5.39
4.90
10.0
3-65
TABLE 3.2.7
PEACH BOTTOM TURBINE TRIP TESTS
REACTIVITY SUMMARY
PEAK REACTIVITY ($ ) TIME OF PEAK (SEC)
CALC. DATA DIFF. CALC. DATA
0.804
0.780
0.836
0;776
0.767
0. 812
3.6%
1. 74
2.54
0.738
0.690
0. 678
0.744
0.696
0.660
3-66
FIGURE 3.2.13
PB TTI CORE AVERAGE POWER
KW
0ll0bJNH
ZK0z
00.0 0.6
lIME (SEC)
FIGURE 3.2.14
PB TT2 CORE AVERAGE PONER
HEASUREO
RETAAN
0M„NHJXK0z
0.0 0.6TIME (SEC)
1.2
FIGURE 3.2.15
PS TT3 CORE AVERAGE PONER
HEASURED
RETRAN
KIU
0Q.
0hlNHJXK0Z
0.0 0.6
nate
(sec)i.2
FIGURE 3.2.16
PB TTi LEVEL A AVERAGE LPRM
HEASURED
RHHAN
0hlNHJXK0Z
0.0 0.6TDIE (SEC)
FIGURE 3.2.17
PB TTI LEVEL B AVERAGE LPRM
0Id„
HJZK0Z
0.0 0.6TIME (SEC)
FIGURE 3.2.18
PB TTi LEVEL C AVERAGE LPRM
KLU
0Q.
0hlNHJXK0z
0.0 0.6TINE (SEC)
FIGURE 3.2.19
PS TTi LEYEI 0 AYERAGE ARM
0WN
ZK0z
0.0 0.6TIME (SEC)
FIGURE 3.2.20
PB TT2 LEVEL A AVERAGE LPRM
8SUAED
p AHRN
0ld
JZK0z
0.0 0.6nHE (sEc)
i.2
FIGURE 3.2.21
PB TT2 LEVEI B AVERAGE ARM
REASUREDC
RETRAN
KIIJ
0Q.
0tdNHJZK0z
0.0 0.6TINE (SEC)
1.2
FIGURE 3.2.22
PB TT2 LEVEL C AVERAGE LPRM
0tU„N
JXK0Z
0.0 0.6TIME (SEC)
FIGURE 3.2.23
PB TT2 LEVEL D AVERAGE LPRM
0hlN
JXK0z
00.0 D.6
, TIME (SEC)
FIGURE 3.2.24
PB TT3 LEVEL A AVERAGE LPRM
ntd
HJXK0Z
0.0 0.6TINE (SEC)
FXGURE 3.2.25
PB TT3 LEVEL B AVERAGE LPRM
ntdNH
XK0Z
0.0 0.6TIHE (SEC)
FIGURE 3.2.26
PB TT3'LEVEL C AVERAGE LPRM
0lU
JXK0Z
0.0 0.6TINE (SEC)
FIGURE 3.2.27
PB TT3 LEVEL D AVERAGE ARM
HEASURED
RETAN
KIUg0
0lU„N
JXK0z
0.0 0.6TINE (SEC)
1.2
FIGURE 3.2.28
O
PB TTl REACTIVITY
HEASUAED
RET. TOTlL6RA
Cl
0
I-HQ)oH
0IUK
tll
OI
N
0I 0.0 0,6
TIME (SEC)f.2
FIGURE 3.2.29
PB TT2 REACTIVITY
HEASURED
RET. TOTaL
ERAH
l0
0
HN>oHI"0
UJ
KOl
OI
8OI 0.0 0.6
TIME (SEC)
FIGURE 3.2.30
0PB TT3 REACTIYITY
HEASURED
BET. TOTAL
a0
0-I-Hol>oI"0
IIJK
Ol
0I
0O
0.0 0.6Trwe (sec)
4. 0 LICENSING BASIS ANALYSIS
A broad spectrum of transient events have been analyzed for WNP-2;
the results are presented in the Final Safety Analysis Report.
These events cover a wide range of scenarios and conditions
contributing to Technical Specification Limits. Most of these
transient events are not sensitive to changes in reload core
configuration, or are within the conservative limits established by
the original FSAR analy'sis. Changes in fuel design and core
configuration are usually bounded by the analysis of selected
limiting events. Based on previous analyses performed by vendors
for WNP-2 + and utilities on similar plants , the two most
limiting events requiring a reanalysis with each reload core are:
1. Load Rejection Without Bypass (LRNB)
2. Feedwater Controller Failure to Maximum Demand (FWCF)
The results of these transients generally determine Technical
Specification limits for minimum critical power ratio (MCPR). This
chapter describes the system analysis for these transients. The
full licensing FSAR analysis, the sensitivity analysis and the hot
channel analysis from which the operating limits are obtained willbe described in a separate applications topical report (to be
submitted later). The purpose of including this section is merely
to indicate the parameters that typically change in modelling
conservative FSAR transients and to show the WNP-2 model, with theparameters modified, indeed calculates conservative results.
4-1
4.1 Licensing Basis Model
The licensing basis model described in this chapter is a generic
model using the Cycle 4 core configuratiop. For future applica-tions, specific reload configurations and plant parameters will be
used. These calculations are typical of planned WNP-2 reload
analyses.
The licensing basis RETRAN model is 'a modification of the WNP-2
best-estimate model. The modifications assure the cons'ervatism ofthe calculated results by using the values of the key parameters
which bound the expected operating range.
Table 4.1 compares licensing basis model. inputs with the nominal
values. The nominal values and conditions show conservatism in the
licensing basis modeling.
The licensing analysis performed in this report uses the end-of-cy-
cle exposure for the calculation of the nuclear design data. At
EOC, control rods are withdrawn from the core to counteract the
consumption of excess reactivity. The average control rod scram
distanc'e is greater with more rods withdrawn, so scram performance
degrades near the end of cycle. Scram reactivity insertion rate isthe dominant power reversal phenomenon for pressurization tran-sients; the most severe results occur at the maximum cycle
exposure, when scram performance is least effective.
4-2
l1
TABLE 4.1
INPUT PARAMETERS AND INITIALTRANSIENT CONDITIONSCOMPARISON OF LICENSING BASIS AND NOMINAL PLANT CONDITIONS
Parameter Nominal Licensin Basis
Core Exposure
Thermal Power (MWt)
Steam Flow (lbs/sec)
BOC — EOC
3323
3970.97
Feedwater Flow Rate (lbs/sec) 3970.97
EOC
3468
4161.11
4161.11
Feedwater Temperature ('F)
Vessel Dome Pressure (psia)
Rod Insertion Speed
Core Inlet Enthalpy (Btu/lb)Fuel Rod Gap Conductance
420'020
Measured
527.6
AxiallyNon-uniform
424
1035
Tech. Spec.
529.3
Uniform
Jet Pump Ratio 2.33
Safety/Relief Valves Relief Function (psig)
Fuel Radial Heat Generation Non-uniform Uniform
2.41
Group 1Group 1Group 2Group 2Group 3.Group 3Group 4Group 4Group 5Group 5OpeningClosingOpening
Opening SetpointClosing SetpointOpening SetpointClosing SetpointOpening SetpointClosing SetpointOpening SetpointClosing SetpointOpening SetpointClosing Setpoint,Stroke Time (sec)Stroke Time (sec)Delay Time (sec)
10761026108610361096104611061056111610660.070.00.3
11061056111610661126107611361086114610960.10.00.4
a. RETRAN will adjust this value at initialization to completethe heat balance.
4-3
TABLE 4.1
INPUT PAEVLHETERS AND INITIALTRANSIENT CONDITIONSCOMPARISON OF LICENSING BASIS AND NOMINA) PLANT CONDITIONS
(Continued)
Parameter Nominal Licensin Basis
Safety/Relief Valves Safety Function (psig)Group 1Group 1Group 2Group 2Group 3Group 3Group 4Group 4Group 5Group 5OpeningClosingOpening
Opening SetpointClosing SetpointOpening SetpointClosing SetpointOpening SetpointClosing SetpointOpening SetpointClosing SetpointOpening SetpointClosing SetpointStroke Time (sec)Stroke Time (sec)Delay Time (sec)
1150112611751151
~ . 1185116111951171120511810.070.00.3
11771153118711631197117312071183121711930.10.00.4
Reactor Protection System
High Flux Scram, 4 NBR 118
High Vessel Dome Pressure 1037Scram (psig)
APRM Thermal Trip (4 NBR 113.5at 1004 Core Flow)
Low, Water Level (L3), in 13above instrument zero
Turbine Stop Valve Closure 5Position Scram (4 Closed)
MSIV Closure Position Scram 10(4 Closed)
126 2
1071
122. 03
7.5
10
15
4 4
TABLE 4.1
INPUT PAEQMETERS AND INITIALTRANSIENT CONDITIONSCOMPARISON OF LICENSING BASIS AND NOMINAL PLANT CONDITIONS
(Continued)
Parameter
Containment Isolation and Pump TripLow Water Level (L2), inbelow instrument zero
Low Pressure in Steamline(psig)
RPT High Vessel Pressure(psig)
RPT Delay Time (msec)
High Water Level — Turbine andFeedwaters Pump Trip (inchesabove instrument zero)
Recirculation Pump Moment ofInertia (10~ ibm — ft~)
Nominal
50
831
" 1135
97
54.5
2.27
Licensin Basis
70
795
1170
190
59.5
2.47
I 4-5
The analyses. in this report used the technical specification limitson control rod movement versus time. Table 4.2 shows the assumed
rod motion following scram+. Actual plant performance data shows
more rapid insertion.
TABLE 4.2Technical Specification Limits
Maximum Control Rod Insertion Time to PositionAfter Deenergization of Pilot Valve Solenoids
Position Inserted fromFull Withdra Notch Number
Time~Sec
6.254'8
'754'7.924'9.584'45)(39)(25)(05)
0.4300.8681.9363.497
The initial steam flow in the licensing basis model is conserva-
tively set at 1054 NBR. For pressurization transients, higher
initial steam flow results in a more rapid pressurization and
higher maximum pressures. The initial power is set to 104.44 NBR
which is consistent with the maximum steam flow of 1054 NBR. The
initial reactor dome pressure is conservatively set to a higher
value of 1035 psia, allowing less analytical margin to the safetylimit.
Unless the problem statepoint requires otherwise, the core flow isinitialized at the rated capacity of 108.5 mlb/hr.
4-6
The RETRAN model uses a fuel rod gap conductance that remains
constant during the transient. The actual gap conductance increases
during power increase transients due to fuel pellet expansion.
Higher gap conductance will lead to faster heat transfer from the
fuel to the coolant, which generates more steam voids. The fasterconversion of fuel stored energy to steam voids in the core helps
to mitigate the transient due to negative void reactivity feedback.
Therefore, the use of a constant gap conductance is conservative
for system analyses involving power increases such as pressuriza-
tion transients.
Conservative equipment design specifications and plant technicalspecification limits are used in the RETRAN model input. These
include relief valve opening response, closure rates for main steam
isolation valves, turbine stop and control valves, reactorprotection system setpoints and delays.
A conservative moment of inertia for the recirculation pump is used
in the licensing basis model. A larger value results in longer
coastdown time after pump trip, delaying the effect of void
formation in the core and increasing the process of void collaps-ing. Positive reactivity effects are magnified by this conserva-
tism.
l 4-7
4.2 Load Rejection Without Bypass (LRNB)
Whenever external disturbances result in loss of electrical load on
the generator, fast closure of the turbine control valves (TCV) isinitiated. The turbine control valves are required to close as
rapidly as possible to minimize overspeed of the turbine generator
rotor. Closure of the main turbine control valves will cause a
sudden reduction in steam flow which results in an increase insystem pressure and reactor shutdown.
4.2.1 Sequence of Events
A loss of generator electrical load at high power with bypass
failure produces the sequence of events listed in Table 4.3.
The closure characteristics of the turbine control valves are
assumed such that the turbine control valves operate in the fullarc (FA) mode and have a full stroke closure time, from fully open
to fully closed, of 0.15 seconds. Sensitivity study" has shown
that the most severe initial condition for this transient is the
assumption of full arc operation at 1054 NBR steam flow. The plantvalue of 0.07 seconds given in Table 4.3 represents actual expected
closure time, since the turbine control valves are partially open
during normal operation.
4-8
TABLE 4.3
Sequence of Events for LRNB Transient
Time-Sec Event
0.0 Turbine generator power load unbalance(PLU) devices trip to initiate turbinecontrol valve fast closure when loss ofelectrical load is detected.
0.0 Turbine bypass valves fail to operate
0.0
0.0
0.07
0 '9
Fast turbine control valve closureinitiates scram tripFast turbine control valve closure ini-tiates a recirculation pump trip (RPT)
Turbine control valves closed
Recirculation pump motor/circuit breakersopen, causing decrease in core flow
0.28
1. 35
1. 40
1.44
1.50
1.63
4.43
5.0
Control rod insertion starts (scram tripsignal at 0 sec, RPS delay : 0.08 sec,solenoid deenergizing delay : 0.2 sec)
Group 1 relief valves actuated
Group 2 relief valves actuated
Group 3 relief valves actuated
Group 4 relief valves actuated
Group 5 relief valves actuated
Group 5 relief valves close
End of simulation
4-9
4.2.2 Results of LRNB RETRAN Analysis
The WNP-2 LRNB analysis at the end of cycle 4 conditions was
performed with the licensing basis model. Since most. of the fuelin the core at EOC4 was the Advanced Nuclear Fuels (ANF) design,
the average fuel parameters in the best-estimate model were changed
from the GE design to the ANF design. The fast closure of the
turbine control valves (TCV) is simulated by linearly decreasing
the flow at fill Junction 380 (representing steam flow to the
turbine) to zero at 0.07 seconds. Rapid closure of the TCV
initiates a scram.
Several key results of this analysis were compared with analyses ofrecord performed by Advanced Nuclear Fuels (ANF). Zt should be
noted that both sets of analyses were performed conservatively.This comparison is intended to show the similarity of resultsrather than to demonstrate analytical accuracy. The accuracy ofthe WNP-2 RETRAN model is demonstrated by the benchmarks of power
ascension tests reported in Section 3.1.
The pressure in the steam line increases rapidly as shown in Figure
4.2.1. The acronym "LRNB LBM" in the figure stands for Load
Rejection without Bypass Licensing Basis Model. Figure 4.2.2 shows
the vessel steam flow. The flow oscillations are caused by the
pressure wave propagated upstream to the reactor vessel. The
decreased steam flow at about 0.4 seconds causes the rapid
4-10
pressurization of the steam dome and inside the core as shown inFigures 4.2.3, 4.2.4 and 4.2.5. After 0.42 seconds, the net
reactivity becomes positive because the positive void reactivityexceeds the negative scram reactivity. Zt peaks with a value of
approximately 0.76$ at 0.78 seconds then begins to decrease as the
scram reactivity increases, as shown in Figure 4.2.6.
The ANF prediction~~ of dome pressure during the transient is also
shown in Figure 4.2.3. The WNP-, 2 RETRAN model predicts a pressure
which is. higher than that predicted by ANF throughout the tran-sient.
The reactor power, shown in Figure 4.2.7, increases rapidly and
reaches a peak value of 3984 NBR at 0.89 seconds then rapidlydecreases as Doppler feedback and scram reactivity terminate the
power excursion. ANF's prediction of core power is also shown inFigure 4.2.7. The power history predicted by RETRAN peaks earlierin the transient than the ANF prediction at a lower maximum power
level. The earlier power peak can be attributed in part to the
higher pressure throughout the transient. The lower magnitude ofthe peak is likely to be caused by differences in neutronics
calculations leading to differences in kinetics data and cross
sections.
Figure 4.2.8 shows the core average clad surface heat flux. The
initial reduction in clad-to-coolant heat transfer is caused by the
4-11
rise in saturation temperature of the liquid phase due to initialpressure rise in the core. As the power rises, the heat fluxquickly reverses and begins to rise, reaching a peak of 133.44 ofthe rated power value at 1.1 seconds. Following the peak, the heat
flux decreases as the core power decreases with a delay determined
by the fuel rod time constant. ANF's calculation of core average
heat flux is also shown in Figure 4.2.8. The two models predictconsistent trends in heat flux and agree closely in the later partof the transient.
t
The water level and feedwater flow during LRNB are shown in Figures
4.2.9 and 4.2.10. When the TCV fast closure calls for scram, the
feedwater controller reduces the water level setpoint by 18 inches.
It then responds to this setpoint change by reducing feedwater
flow. Pressure variations, steam flow oscillations, and void
collapse contribute to the changing water level throughout. the
remainder of the transient.
Figures 4.2.11 and 4.2.12 give the void fractions at mid-core and
core exit. Core voids collapse as the steamline pressure wave .
reaches the core. For the remainder of the transient, variationsin steam flow and pressure drive oscillations in the void fraction.Figure 4.2.13 shows the recirculation flow. The recirculation pumps
start to coast down after RPT initiation at 0.19 seconds, causing
flow reduction in the core as shown in Figure 4.2.14.
4-12
FIGURE 4.2.1
WNP-2 LRNB LBM — STEAMLINE PRESSURE
T IHE (SEC)
FIGURE 4.2.2
WNP-2 LRNB LBM — VESSEL STEAM FLOW
2TIHE (SEC)
FIGURE 4.2.3
HNP-2 LANB LBM — OOME PRESSURE
+ PACQI - VOL $ 4
)( AMP CALO%ATION
Oa0
llJKDNUltaboo[frs
Q
~I yTIME (SEC)
FIGURE 4.2.4
NNP-2 LRNB LBM — PRESSURE (MID-CORE)
0N
MUlQ
UJKU)Ntuoo0.'eo
Q%I
OO0~I
O 2TIME (SEC)
FIGURE 4.2.5
WNP-2 LRNB LBM — PRESSURE (CORE EXIT)
OOa~I
IdKDQQejo(fPe
g rl
OOO~i e
TIME (SEC)
FIGURE 4.2-6
NNP-2 LRNB LBM — TOTAL REACTIVITY
+ TQZAL REACTIVITY
I-H»+IHlO
WK
TIHE (SEC)
FIGURE 4.2.7
HNP-2 LRNB LBM — CORE POHER
+ CORC PNKA
g ANF CALCAATIOtl
KlU
0Q.o
a
0TIME (SEC)
FIGURE 4.2.8
HNP-2 LRNB LBM — CORE AVERAGE HEAT FLUX
KezR
XoDel
b.
O4O
TIHE (SEC)
FIGURE 4.2.9
OrWNP-2 LRNB LB M — LIQu XO LEVEL
2H
tU
gO~II
0H3UM
O"oTIME [SEC)
FIGURE 4.2.10
WNP-2 LRNB LBM — FEEDWATER FLOW
TIME (SEC)
FIGURE 4.2.11
WNP-2 LRNB LBM — VOID FRAC (MID-COREJ
+ VOIO FRAC VQ..5$
zOHI-
Kh.
0OgQ ~
4OTIME (SEC)
FIGURE 4.2-12
HNP-2 LRNB LBM — VOID FRAC (CORE EXIT)
O
z0HI-O4oglQ.o
0MO
OQ
OoTINE (SEC)
FIGURE 4.2.13
WNP-2 LRNB LBM — RECIRCULATION FLOW
X0<o
OC7 0
TIME (SEC)
FIGURE 4.2.14
WNP-2 LRNB LBN — CORE INLET FLOW
O4
0
ORo
- TIHE (SEC)
4.3 Feedwater Controller Failure to Maximum Demand (FWCF)
This analysis postulates the feedwater controller failure causing
an increase in feedwater flow. The most severe event of this type
is a failure at maximum flow demand. The feedwater controller isforced to its upper limit at the beginning of the event.
4.3.1 Sequence of Events
When feedwater flow exceeds steam flow, the water. level rises.With the feedwater controller disabled, the water level eventually
rises to the high-level reference point, at which time the
feedwater pumps and the main turbine are tripped and a scram isinitiated. Table 4.4 lists the sequence of events for thistransient.
The transient is simulated by programming an upper limit failure inthe feedwater system such that 146 percent feedwater flow occurs atthe design pressure of 1020 psig.
An increase in feedwater flow will cause a corresponding drop infeedwater temperature. However, the relatively large time constant
of the feedwater heaters (order of minutes) plus the flow transporttime (about 10 seconds from heaters to vessel and about 3 seconds
from sparger to core) delay the cold water effects until the event
has run its course and no further consequences are expected.
4-27
Feedwater temperature is therefore assumed to remain constant.
4.3.2 -Results of FWCF RETRAN Analysis
The FWCF event was analyzed at EOC4 conditions using the licensingbasis model. As in the LRNB case, the fuel parameters were changed
to reflect the ANF design.
The feedwater flow is shown in Figure 4.3.1.. The increased feedwa-
ter,. after a transport delay through the lower downcomer and
recirculation loops, causes an increase in the core inlet subcool-
ing as shown in Figure 4.3.2. The reactor water level alsoincreases as the feedwater flow increases as shown in Figure 4.3.3.The water level reaches the high level setpoint at approximately17.7 seconds, causing a turbine trip (Figure 4.3.4). The vessel
pressure begins to increase after the closure of the turbine stop
valves. The turbine bypass valves open (Figure 4.3.5). For highinitial power levels the bypass, which has a design capacity of 254
of rated steam flow, will not divert all of the steam flow and thepressure increases as shown in Figure 4.3.6. As the pressure
increases, the core void collapses. The positive void reactivityfrom core pressurization is initially high enough to overcome the
negative scram reactivity (Figure 4.3.7). A rapid increase inpower begins at 18.1 seconds (Figure 4.3.8), reaching a peak of2454 NBR at 18.6 seconds. Trailing the reactor power, the core
average heat flux peaks at 1244 NBR at 18.8 seconds (Figure 4.3.9).
4-28
The relief valves open after the vessel pressure reaches the
opening setpoints (Figures 4.3.10 through 4.3.14) and additionalsteam is relieved to the suppression pool. The vessel pressure
begins to decrease rapidly. Figure 4.3.15 shows the vessel steam
flow. The fast closure of the turbine stop valves generate pressure
waves which lead to the oscillations in steam flow. The core inletflow and exit flow are shown in Figures 4.3.16 and 4.3.17. Due tothe coastdown of the recirculation pumps, the core flow reduces
near the end of the transient (Figure 4.3.18). Figures 4.3.19 and
4.3.20 show the mid-core and core exit pressures. Figures 4.4,.21
and 4.4.22 show the mid-core and core exit void fractions. The
transient is terminated by the Doppler feedback and the control rod
scram.
The RETRAN results were not compared with ANF analyses of record
for the FNCF transient because different initial conditions were
assumed in the two analyses. The ANF analysis assumed an initialpower level of 474 of rated, while the example calculation reported
here was performed at an initial power level of 104.44 of rated.
4-29
TABLE 4.4
Secpxence of Events for Feedwater Controller Failure
Time-Sec Event
Initiate simulated failure of 146% upper limiton feedwater flow
17.60
17.66
17. 67
L8 vessel Level setpoint trips feedwater pumps
L8 vessel level setpoint trips main turbine.Turbine bypass operation initiatedReactor scram trip actuated from main turbinestop valve position switches (104 closure)
17. 86
19. 12
19. 33
19. 39
19 '719 70
20.56
20.90
21.74
23.49
24.0
Group 2 relief valves
Group 3 relief valves
actuated
actuated
Group 4 relief valves actuated
Group 5 relief valves actuated
Group 5 relief valves start to close
Group 4 relief valves start to close
Group 3 relief valves start to close
Group 2 relief valves start to close
End of simulation
Recirculation pump trip (RPT) actuated by stopvalve position switches (104 closure)
Group 1 relief valves actuated
4-30
FIGURE 4.3.1
WNP-2 FWCF LBM — FEEOWATER FLOW
TINE (SEC)je
FIGURE 4.3.2
HNP-2 FHCF LBM — CORE INLET SUBCOOLING
S
I-Q
NoH00OQ
N
TIME (SEC)f4
FIGURE 4.3.3
WNP-2 FWCF LBM — LIQUIO LEVEL
mg
0HD
H
TIME (SEC)f4
FIGURE 4 ' '
HNP-2 FHCF LBM — TURBINE STEAM FLQN
TIHE (SEC)
FIGURE 4.3.5
WNP-2 FWCF LBM — TURBINE BYPASS FLOW
KSzR
xO
<or
TIME tSEC)
PXGURE 4.3.6
WNP-2 FWCF LBM — DOME PRESSURE
lUKMQ+o0:roQ
OOO~I y
TIHf (SEC)ie
FIGURE 4.3-7
WNP-2 FWCF LBM — TOTAL REACTIVITY
HD>oHI-O
QJK
R
I O
TIME (SEC)
FIGURE 4.3.8
WNP-2 FWCF LBM — CORE POWER
KllJXOQ
0TIHE (SEC)
Ce
FIGURE 4.3.9
WNP-2 FWCF LBM — CORE AVERAGE'HEAT FLUX
. TIME tSEC)i4
FIGURE 4.3-10
NNP-2 FHCF LBM — GROUP l. SRV FLOW
. TIHE (SEC)
FIGURE 4.3.11
WNP-2 FWCF LBM — GROUP 2 SRV FLOW
30<o
TIME (SEC)
FIGURE 4.3.12
WNP-2 FWCF LBM — GROUP 3 SRV FLOW
04o
TIME (SEC)
FIGURE 4.3.13
WNP-2 FWCF LBM — GROUP 4 SRV FLOW
TIME (SEC)
FIGURE 4 ' '4
WNP-2 FWCF LBM — GROUP 5 SRV FLOW
TIHE (SEC)
FIGURE 4.3.15
WNP-2 FWCF LBM — VESSEL STEAM FLOW
ALON ~ SfO
. TIHE (SEC)
FIGURE 4.3.16
WNP-2 FWCF LBM — CORE INLET FLOW
TIHE (SEC)14
FIGURE 4.3.17
O~I
WNP-2 FWCF LBM — CORE EXIT FLOW
mz1C
'0
0h.
O4O
TIHE (SEC)$4
FIGURE 4.3.18
WNP-2 FWCF LBM — RECIRCULATION FLOW
Kmz
TAHE (SEC)
FIGURE 4.3.19
WNP-2 FWCF LBM — PRESSURE tMED-CORE)
OO0
tlJKlO0)~OQPoQ
OA
O
'IME (SEC)
FIGURE 4.3.20
NNP-2 FHCF LBH — PRESSURE (CORE EXIT)
HMQ
lUKD0)MQJ
0gasg tl
OOe~I 0
TIHE (SEC)
FIGURE 4.3.21
WNP-2 FWCF LBM — VOID FRAC (MID-CORE)
zOHIO
Kh.
0H0
aOO
TINE (SEC)
FXGURE 4.3.22
WNP-2 FWCF LBM — VOID FRAC (CORE EXIT)
Z0HI-O
Kh.
0H0
O0O
O
TIME (SEC)
4.4 Summary of Transient Analysis
The key transient simulation„. results for the two MCPR limitingtransients as calculated by RETRAN are summarized in Table 4.5.
TABLE 4.5
Summary of Thermal-Limiting Transient Results
LRNB FWCF
Max Power (%NBR)
Time at'ax power (seconds)
Max core avg heat flux (4NBR)
Time at max heat flux (seconds)
Max dome pressure (psia)
Time at max dome pressure (sec)
398
0.89
133
1207
1 ~ 9
245
18.6
124
18.8
1175
19.5
4-53
II
I
I
I
5.0 SUMMARY AND CONCLUSIONS
Benchmark analyses covering specific Power Ascension Tests as
described in Section 3.1 demonstrate the capability of the WNP-2
RETRAN model to predict core and system behavior during normal
operation and mild transients. These analyses validate the
modeling of the feedwater and pressure regulator control systems
and the performance of the recirculation pumps, jet pumps, and
steam lines as modeled for,WNP-2.
Benchmark analyses covering the turbine trip tests performed atPeach Bottom 2 at the end of Cycle 2 as described in Section 3.2
demonstrate RETRAN's ability to model conditions more challengingthan the WNP-2 startup tests and the Supply System technicalstaff's competence to perform these analyses. These analyses
validate the capabilities of the modeling beyond the normal
operating envelope of the reactor.
Example calculations covering typical limiting transients as
reported in Chapter 4 demonstrate the WNP-2 RETRAN model's
ability to predict system performance under conditions which
challenge operating limits. These analyses show consistency withexisting analyses of record and validate the Supply. System
technical staff's ability to formulate and analyze limitingtransient events.
5-1
The analyses performed in this report demonstrate the ability ofthe WNP-2 RETRAN model and the qualifications of the Supply
System technical staff to predict the course of a wide variety oftransient events. The model is applicable to the evaluation ofnormal- and anticipated operation for plant operational support
and core reload analysis.
5-2
6.0 REFERENCES
1 ~
2 ~
J.H." McFadden et al., "RETRAN-02 — A Program for TransientThermal-Hydraulic Analysis of Complex Fluid Flow Systems,"EPRI NP-1850-CCM-A, Revision 4, Volumes I-III, Electric PowerResearch Institute, November 1988.
B.M. Moore, A.G. Gibbs, J.D. Imel, J.D. Teachman, D.H.Thomsen, and W.C. Wolkenhauer, "Qualification of Core PhysicsMethods for BWR Design and Analysis," WPPSS-FTS-127,Washington Public Power Supply System, March 1990.
3. J.A. McClure et al., "SIMTRAN-E — A SIMULATE-E to RETRAN-02Data Link," EPRI NP-5509-CCM, Electric Power Research Insti-tute, December 1987.
4. C.W. Stewart et al.,"VIPRE-01 — A Thermal-Hydraulic Code forReactor Cores," EPRI NP-254-CCM-A, Revision 3, Volumes I-III,Electric Power Research Institute, August 1989.
D.L. Hagerman, G.A. Reymann, and R.E. Manson, "MATPROVersion 11 (Revision 2): A Handbook of Materials Propertiesfor Use in, the Analysis of Light Water Reactor Fuel RodBehavior," NUREG/CR-0479, TREE-1280, Revision 2, IdahoNational Engineering Laboratory, August 1981.
6. "WREM, Water Reactor Evaluation Model, Revision 1," NUREG-75/065, U.S. Nuclear Regulatory Commission, May 1975.
7. WPPSS Nuclear Plant 2 Updated Final Safety Analysis Report,Washington Public Power Supply System, 1989.
8.
9.
"Qualification of the One-Dimensional Core Transient Modelfor Boiling Water Reactors," NEDO-24154, Volume 1, GeneralElectric Company, October 1978.
Letter, '. Armenta (GE) to W.C. 'olkenhauer (WPPSS),"Instruction for Use of 4-Quadrant Curve," dated March 26,1985.
10. "Recirculation System Performance," Publication 457HA802,General Electric Company, September 1976.
11. B.J. Gitnick et al., "FIBWR — A Steady-State Core FlowDistribution Code for Boiling Water Reactors," EPRI NP-1924-CCM, Electric Power Research Institute, July 1981.
12. R.E. Polomik and S.T. Chow, "Hanford-2 Nuclear Power StationControl System Design Report," GEZ-6894, General ElectricCompany, February 1980.
6-1
13 ~
14.
15.
16.
17.
18.
19.
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